Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 61655 

This notice is distributed by mail to 

several hundred subscribers; if you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301) 415–1969. 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to dkw@nrc.gov. 

Dated: October 20, 2005. 

R. Michelle Schroll, 

Office of the Secretary. 

[FR Doc. 05–21337 Filed 10–21–05; 9:54 am] 

BILLING CODE 7590–01–M 

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a.(2) of the 

Atomic Energy Act of 1954, as amended 

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

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 September 

30 to October 13, 2005. The last 

biweekly notice was published on 

October 11, 2005 (70 FR 59082). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. Within 60 days after the 

date of publication of this notice, the 

licensee may file a request for a hearing 

with respect to issuance of the 

amendment to the subject facility 

operating license and any person whose 

interest may be affected by this 

proceeding and who wishes to 

participate as a party in the proceeding 

must file a written request for a hearing 

and a petition for leave to intervene. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules and 

Directives Branch, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be delivered to Room 6D22, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

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

Copies of written comments received 

may be examined at the Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. The filing of 

requests for a hearing and petitions for 

leave to intervene is discussed below. 

Within 60 days after the date of 

publication of this notice, the licensee 

may file a request for a hearing with 

respect to issuance of the amendment to 

the subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

intervene. Requests for a hearing and a 

petition for leave to intervene shall be 

filed in accordance with the 

Commission’s ‘‘Rules of Practice for 

Domestic Licensing Proceedings’’ in 10 

CFR part 2. Interested persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

61656 Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

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. 

Dominion Nuclear Connecticut Inc., et 

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

Station, Unit No. 3, New London 

County, Connecticut 

Date of amendment request: 

September 13, 2005. 

Description of amendment request: 

The amendment would reduce the 

temperature at which shutdown and 

control rod cluster control assemblies 

(RCCA) drop testing is done from greater 

than or equal to 551 °Fahrenheit (F) to 

greater than or equal to 500 °F. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

Criterion 1: Does the proposed 

amendment involve a significant 

increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. DNC [Dominion 

Nuclear Connecticut, Inc.] is proposing 

to change the temperature at which the 

shutdown and control RCCA drop tests 

are performed from ‘‘greater than or 

equal to 551 °F,’’ to ‘‘greater than or 

equal to 500 °F.’’ The proposed change 

does not modify any plant equipment 

and does not impact any failure modes 

that could lead to an accident. 

Additionally, the proposed change has 

no effect on the consequence of any 

analyzed accident since the change does 

not affect the function of any equipment 

credited for accident mitigation. Based 

on this discussion, the proposed 

amendment does not increase the 

probability or consequences of an 

accident previously evaluated. 

Criterion 2: Does the proposed 

amendment create the possibility of a 

new or different kind of accident from 

any accident previously evaluated? 

Response: No. The proposed change 

does not modify any plant equipment 

and there is no impact on the capability 

of existing equipment to perform its 

intended functions. No system setpoints 

are being modified and no changes are 

being made to the method in which 

plant operations are conducted. No new 

failure modes are introduced by the 

proposed change. The proposed 

amendment does not introduce accident 

initiators or malfunctions that would 

cause a new or different kind of 

accident. 

As noted above, the proposed change 

does not affect the revisions to plant 

procedures, which were made to 

address Westinghouse Nuclear Safety 

Advisory Letter, NSAL–00–016 (Rod 

Withdrawal from Subcritical Protection 

in Lower Modes, issued in 2000). 

Therefore, the proposed amendment 

does not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

Criterion 3: Does the proposed 

amendment involve a significant 

reduction in a margin of safety? 

Response: No. The TS [technical 

specification] change does not involve a 

significant reduction in margin because 

the acceptance criterion for the RCCA 

drop time will not change. The 

proposed change will reduce the 

minimum RCCA drop test temperature 

from greater than or equal to 551 °F to 

greater than or equal to 500 °F. This will 

slightly increase the measured test 

Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 61657 

RCCA drop time. However, the 

measured test RCCA drop time is 

required to remain within the current 

TS limit of 2.7 seconds and the 2.19 

seconds for surveillance testing 

acceptance criteria (plant specific 

seismic allowance of 0.51 seconds). The 

proposed change does not affect any of 

the assumptions used in the accident 

analysis, nor does it affect any 

operability requirements for equipment 

important to plant safety. Therefore, the 

margin of safety is not impacted by the 

proposed amendment. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Lillian M. 

Cuoco, Senior Nuclear Counsel, 

Dominion Nuclear Connecticut, Inc., 

Waterford, CT 06141–5127. 

NRC Section Chief: Darrell J. Roberts. 

Dominion Nuclear Connecticut Inc., et 

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

Station, Unit No. 3, New London 

County, Connecticut 

Date of amendment request: 

September 13, 2005. 

Description of amendment request: 

The changes revise surveillance 

requirements for the recirculation spray 

system (RSS) to verify proper initiation 

of recirculation spray through actuation 

by the refueling water storage tank 

(RWST) low-low level signal instead of 

actuation by a timer. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

Criterion 1: Does the proposed 

amendment involve a significant 

increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. The RSS is only an 

accident mitigation system. As such, 

changes in the operation of RSS cannot 

have an impact on the probability of an 

accident. The delay in the start of the 

RSS pump is to assure there is sufficient 

water in the containment sump for 

adequate RSS pump NPSH [net positive 

suction head] and margin to suction 

pipe flashing in light of the debris 

analysis conducted in response to GL 

[Generic Letter] 2004–02. Containment 

analyses have been performed to 

demonstrate that there is no impact on 

the peak containment pressure and 

temperature following a LOCA [loss-ofcoolant accident]. While there are some 

changes in the predicted post-LOCA 

environmental conditions, evaluations 

have been performed to show that there 

is no significant impact on the 

environmental qualification for 

equipment inside containment. The 

impact to piping and supports has been 

demonstrated to be acceptable without 

modification. Delay in RSS spray start 

will result in a reduction in diesel 

generator loading since the RSS pumps 

and the RHS pumps will no longer be 

running concurrently. The reduction in 

iodine removal efficiency during the 

delay period is more than offset by 

elimination of over-conservatisms in 

assumptions for long term iodine 

removal by the RSS system. The net 

impact is a reduction in the predicted 

offsite doses and control room doses 

following a design basis LOCA. Based 

on this discussion, the proposed 

amendment does not increase the 

probability or consequence of an 

accident previously evaluated. 

Criterion 2: Does the proposed 

amendment create the possibility of a 

new or different kind of accident from 

any accident previously evaluated? 

Response: No. The proposed 

modification alters the RSS pump 

circuitry by initiating the start sequence 

with an existing RWST low-low level 

signal instead of a timer. The timer is 

now used to sequence pump starts. The 

pump function is not changed in any 

way. The proposed amendment does not 

introduce failure modes, accident 

initiators, or malfunctions that would 

cause a new or different kind of 

accident. Therefore, the proposed 

amendment does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Criterion 3: Does the proposed 

amendment involve a significant 

reduction in a margin of safety? 

Response: No. The proposed change 

ensures that adequate margin to suction 

line flashing and NPSH margin exists 

for proper operation of the RSS pumps 

once the effects of debris are considered 

as required per GL 2004–02. Function of 

the pumps is not affected. Analyses 

have been performed that show the 

containment design basis limits are 

satisfied and the post-LOCA offsite and 

control room doses meet the required 

criteria. Therefore, based on the above, 

the proposed amendment does not 

involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Lillian M. 

Cuoco, Senior Nuclear Counsel, 

Dominion Nuclear Connecticut, Inc., 

Waterford, CT 06141–5127. 

NRC Section Chief: Darrell J. Roberts. 

Duke Energy Corporation, et al., Docket 

Nos. 50–369 and 50–370, McGuire 

Nuclear Station, Units 1 and 2, 

Mecklenburg County, North Carolina 

and Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of amendment request: June 29, 

2005. 

Description of amendment request: 

The amendment would revise Technical 

Specification Bases Section 3.6.11, ‘‘Air 

Return System (ARS),’’ and the Updated 

Final Safety Analysis Reports (UFSAR), 

Section 6.2, ‘‘Containment Systems,’’ for 

McGuire Nuclear Station, Units 1 and 2 

and Catawba Nuclear Station, Units 1 

and 2. The licensee proposes to 

implement an additional manual 

operator action to respond to NRC 

Bulletin 2003–01, ‘‘Potential Impact of 

Debris Blockage on Emergency Sump 

Recirculation at Pressurized-Water 

Reactors.’’ This amendment would 

allow plant operators to manually start 

one air return fan at a containment 

pressure of 1 psig prior to the automatic 

9 minutes (+ 1 minute) delayed start 

described in the UFSAR. 

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: 

First Standard 

Does the change involve a significant 

increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The manual start of an Air Return 

System (ARS) fan will not result in a 

significant increase in the probability of 

an accident previously evaluated. The 

starting of an ARS fan is not considered 

to be an initiator of any accident or 

transient. This action is not taken 

during normal plant operation, but in 

response to an accident. The ARS fans 

do not operate to provide any normal 

ventilation requirement. The 

Containment Pressure Control System 

(CPCS) is provided to prevent excessive 

depressurization of the containment 

through inadvertent or excessive 

operation of certain engineered safety 

features. The CPCS prevents the 

61658 Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 

inadvertent actuation of an ARS fan 

during normal operation. 

This change is being requested in 

order to mitigate the consequences of a 

small break loss of coolant accident 

(SBLOCA) and help prevent or delay 

reaching the initiation pressure setpoint 

for containment spray, thereby reducing 

associated problems with possible sump 

debris buildup. SBLOCA events are 

bounded by the consequences of a 

design basis large break [loss of coolant 

accident] LOCA as addressed in Section 

15 of the McGuire and Catawba 

[Updated Final Safety Analysis Report] 

UFSARs. Accordingly, this amendment 

will not involve a significant increase in 

the consequences of an accident 

previously evaluated. 

Second Standard 

Does the change create the possibility 

of a new or different kind of accident 

from any accident previously evaluated? 

No. The change proposed in this 

[license amendment request] LAR does 

not involve a physical alteration to the 

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

equipment will be installed) or a change 

in the methods governing any normal 

plant operation. It does allow for the 

early start of one ARS fan during a 

SBLOCA event with containment 

pressure greater than 1 psig and less 

than 3 psig. This change will not affect 

or degrade the ability of the ARS to 

perform its specified safety functions. 

Accidents of a different type are 

credible accidents that the proposed 

amendment could create that are not 

bounded by UFSAR evaluated 

accidents. This amendment allows for 

the manual start of an ARS fan 

following a SBLOCA within the 

containment. No new failure modes are 

introduced due to the manual start of an 

ARS fan. The circuit used to manually 

start an ARS fan does not interfere with 

the automatic signal to start an ARS fan. 

This change does not require any 

modifications to the control circuitry for 

the ARS. The starting of an ARS fan is 

not considered to be an initiator of any 

accident or transient. This action 

(starting of an ARS fan) is not taken 

during normal operation, but in 

response to an accident. Previous 

accidents considered incredible are not 

made more likely by this change. A 

human performance error, such as 

starting the ARS fan too early, too late, 

or not at all, would not result in a 

substantial difference in the calculated 

differential pressure across the divider 

deck. Since no new malfunctions of 

equipment with a different result are 

introduced, all effects of any 

malfunctions are bounded by those 

already evaluated in the UFSAR. Thus 

it is concluded that the change 

contained in this LAR will not create 

the possibility of a new or different kind 

of accident from any accident 

previously evaluated. 

Third Standard 

Does the proposed change involve a 

significant reduction in a margin of 

safety? 

No. The early manual start of an ARS 

fan for SBLOCA events will not reduce 

the ability of this system to perform its 

design functions to assure the rapid 

return of air from the upper to the lower 

containment compartment after the 

initial blowdown following a Design 

Basis Accident (DBA). The return of this 

air to the lower compartment and 

subsequent recirculation back up 

through the ice condenser assists in 

cooling the containment atmosphere 

and limiting post accident pressure and 

temperature in containment to less than 

design values. Limiting pressure and 

temperature also reduces the release of 

fission product radioactivity from 

containment to the environment in the 

event of a DBA. Therefore, there are no 

adverse dose effects from the early start 

of the ARS fan or from the delay of 

containment spray based on the current 

licensing basis. 

Analyses have shown that there will 

be no fan or damper malfunction due to 

the early manual start of a fan. The other 

functions of the system are not affected 

by the change proposed in this LAR. 

The manual start of the ARS during a 

SBLOCA will help maintain the margin 

of safety by forcing air and steam 

through the ice condenser with a 

subsequent reduction in the rate of 

pressure increase in the containment, 

and a delay in reaching the actuation 

setpoint for the containment spray 

system. The containment spray system 

will continue to be initiated at the 

normal setpoint pressure of the system 

(-3 psig). Therefore, the proposed 

changes listed above 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: Ms. Lisa F. 

Vaughn, Legal Department (PB05E), 

Duke Energy Corporation, 422 South 

Church Street, Charlotte, North Carolina 

28201–1006. 

NRC Section Chief: Evangelos C. 

Marinos. 

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 amendment request: August 

17, 2005. 

Description of amendment request: 

The proposed changes will revise the 

Operating License Condition 2.C.(41), 

Fire Protection Program, to add a 

reference to the Nuclear Regulatory 

Commission (NRC) safety evaluation 

that allows the application of National 

Fire Protection Agency risk-informed, 

performance based fire protection 

methods and tools that have been 

approved by the NRC. 

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 activity 

involves the use of a risk-informed, 

performance-based method to identify 

those circuits where a single fire could 

damage more than one safe shutdown 

train. These circuits would then be 

provided with one hour rated fire wrap. 

With the exception of the fire wrap 

itself, the proposed activity does not 

result in any physical changes to safetyrelated structures, systems, or 

components (SSCs), or the manner in 

which safety-related SSCs are operated, 

maintained, modified, tested, or 

inspected. The proposed activity does 

not degrade the performance or increase 

the challenges of any safety-related 

SSCs assumed to function in the 

accident analysis. As a result, the 

proposed activity does not introduce 

any new accident initiators. In addition, 

fires are not an accident that is 

previously evaluated in Chapter 15. 

Regardless, the proposed activity does 

not change the probability of a fire 

occurring since fire ignition frequency is 

independent of the presence of the fire 

wrap. The consequences of the 

proposed activity are bounded by the 

fire safe shutdown analysis, which 

assumes one train is free of fire damage. 

Therefore, providing one hour rated 

fire wrap for those circuits where a 

single fire could damage more than one 

safe shutdown train does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 61659 

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 activity 

involves the use of a risk-informed, 

performance-based method to identify 

those circuits where a single fire could 

damage more than one safe shutdown 

train. These circuits would then be 

provided with one hour rated fire wrap. 

With the exception of the fire wrap 

itself, the proposed activity does not 

result in any physical changes to safetyrelated structures, systems, or 

components (SSCs), or the manner in 

which safety-related SSCs are operated, 

maintained, modified, tested, or 

inspected. The proposed activity does 

not degrade the performance or increase 

the challenges of any safety-related 

SSCs assumed to function in the 

accident analysis. As a result, the 

proposed activity does not introduce 

nor increase the number of failure 

mechanisms of a new or different type 

than those previously evaluated. The 

fire safe shutdown analysis assumes one 

train is maintained free of fire damage. 

Therefore, providing one hour rated 

fire wrap for those circuits where a 

single fire could damage more than one 

safe shutdown train 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 activity 

involves the use of a risk-informed, 

performance-based method to identify 

those circuits where a single fire could 

damage more than one safe shutdown 

train. These circuits would then be 

provided with one hour rated fire wrap. 

With the exception of the fire wrap 

itself, the proposed activity does not 

result in any physical changes to safetyrelated structures, systems, or 

components (SSCs), or the manner in 

which safety-related SSCs are operated, 

maintained, modified, tested, or 

inspected. The proposed activity does 

not degrade the performance or increase 

the challenges of any safety-related 

SSCs assumed to function in the 

accident analysis. 

The proposed activity does not impact 

plant safety since the conclusions of the 

fire safe shutdown analysis remain 

unchanged. 

Therefore, providing one hour rated 

fire wrap for those circuits where a 

single fire could damage more than one 

safe shutdown train 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: Nicholas S. 

Reynolds, Esquire, Winston and Strawn, 

1700 K Street, NW., Washington, DC 

20006–3817. 

NRC Section Chief: David Terao. 

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: July 21, 

2005. 

Description of amendment request: 

The amendment proposes to replace the 

existing steam generator tube 

surveillance program with that being 

proposed by the Technical Specification 

Task Force (TSTF) in TSTF 449, 

Revision 4. 

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 

requires a Steam Generator Program that 

includes performance criteria that will 

provide reasonable assurance that the 

steam generator (SG) tubing will retain 

integrity over the full range of operating 

conditions (including startup, operation 

in the power range, hot standby, 

cooldown and all anticipated transients 

included in the design specification). 

The SG performance criteria are based 

on tube structural integrity, accident 

induced leakage, and operational 

leakage. 

The structural integrity performance 

criterion is: 

Structural integrity performance 

criterion: All in-service steam generator 

tubes shall retain structural integrity 

over the full range of normal operating 

conditions (including startup, operation 

in the power range, hot standby, and 

cool down and all anticipated transients 

included in the design specification) 

and design basis accidents. This 

includes retaining a safety factor of 3.0 

against burst under normal steady state 

full power operation primary to 

secondary pressure differential and a 

safety factor of 1.4 against burst applied 

to the design basis accident primary to 

secondary pressure differentials. Apart 

from the above requirements, additional 

loading conditions associated with the 

design basis accidents, or combination 

of accidents in accordance with the 

design and licensing basis, shall also be 

evaluated to determine if the associated 

loads contribute significantly to burst or 

collapse. In the assessment of tube 

integrity, those loads that do 

significantly affect burst or collapse 

shall be determined and assessed in 

combination with the loads due to 

pressure with a safety factor of 1.2 on 

the combined primary loads and 1.0 on 

axial secondary loads. 

The accident induced leakage 

performance criterion is: The primary to 

secondary accident induced leakage rate 

for any design basis accidents, other 

than a SG tube rupture, shall not exceed 

the leakage rate assumed in the accident 

analysis in terms of total leakage rate for 

all SGs and leakage rate for an 

individual SG. Leakage is not to exceed 

540 gallons per day through any one SG, 

except for specific types of degradation 

at specific locations as described in 

paragraph c of the Steam Generator 

Program. 

The operational leakage performance 

criterion is: The RCS operational 

primary to secondary leakage through 

any one SG shall be limited to ≤ 75 

gallons per day per SG. 

A steam generator tube rupture 

(SGTR) event is one of the design basis 

accidents that is analyzed as part of a 

plant’s licensing basis. In the analysis of 

a SGTR event, a bounding primary to 

secondary leakage rate equal to the 

leakage rate associated with a doubleended rupture of a single tube is 

assumed. 

For other design basis accidents such 

as main steam line break (MSLB), 

control element assembly (CEA) 

ejection, and reactor coolant pump 

seized rotor/sheared shaft, the tubes are 

assumed to retain their structural 

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

rupture). The accident induced leakage 

criterion introduced by the proposed 

changes account for tubes that may leak 

during design basis accidents. The 

accident induced leakage criterion 

limits this leakage to no more than the 

value assumed in the accident analysis. 

The SG performance criteria proposed 

change identify the standards against 

which tube integrity is to be measured. 

Meeting the performance criteria 

provides reasonable assurance that the 

SG tubing will remain capable of 

fulfilling its specific safety function of 

maintaining reactor coolant pressure 

boundary integrity throughout each 

operating cycle and in the unlikely 

event of a design basis accident. The 

performance criteria are only a part of 

61660 Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 

the Steam Generator Program required 

by the proposed change. The program, 

defined by NEI [Nuclear Energy 

Institute] 97–06, Steam Generator 

Program Guidelines, includes a 

framework that incorporates a balance 

of prevention, inspection, evaluation, 

repair, and leakage monitoring. 

The consequences of design basis 

accidents are, in part, functions of the 

Specific Activity in the primary coolant 

and the primary to secondary leakage 

rates resulting from an accident. 

Therefore, limits are included in the 

plant technical specifications for 

operational leakage and for Specific 

Activity in primary coolant to ensure 

the plant is operated within its analyzed 

condition. For those analyzed events 

that do not result in faulted steam 

generators, greater than or equal to 75 

gpd [gallons per day] primary to 

secondary leakage per steam generator is 

assumed in the analysis. For those 

analyzed events that result in a faulted 

steam generator (e.g., MSLB), 540 gpd 

primary to secondary leakage is 

assumed though the faulted steam 

generator while greater than or equal to 

75 gpd primary to secondary leakage is 

assumed though the intact steam 

generator. 

The proposed change does not affect 

the design of the SGs, their method of 

operation, or primary coolant chemistry 

controls. The proposed approach 

updates the current technical 

specifications and enhances the 

requirements for SG inspections. The 

proposed change does not adversely 

impact any other previously evaluated 

design basis accident and is an 

improvement over the current Technical 

Specifications. 

Therefore, the proposed change does 

not affect the consequences of a SGTR 

accident and the probability of such an 

accident is reduced. In addition, the 

proposed changes do not affect the 

consequences of other design basis 

events. 

2. Does the proposed change create 

the possibility of a new or different kind 

of accident from any previously 

evaluated? 

Response: No. The proposed 

performance based requirements are an 

improvement over the requirements 

imposed by the current technical 

specifications. 

Implementation of the proposed 

Steam Generator Program will not 

introduce any adverse changes to the 

plant design basis or postulated 

accidents resulting from potential tube 

degradation. The result of the 

implementation of the Steam Generator 

Program will be an enhancement of SG 

tube performance. Primary to secondary 

leakage that may be experienced during 

all plant conditions will be monitored to 

ensure it remains within current 

accident analysis assumptions. 

The proposed change does not affect 

the design of the SGs, their method of 

operation, or primary or secondary 

coolant chemistry controls. In addition, 

the proposed change does not impact 

any other plant system or component. 

The change enhances SG inspection 

requirements. Therefore, the proposed 

change does not create the possibility of 

a new or different type of accident from 

any accident previously evaluated. 

3. Does the proposed change involve 

a significant reduction in a margin of 

safety? 

Response: No. The SG tubes in 

pressurized water reactors are an 

integral part of the reactor coolant 

pressure boundary and, as such, are 

relied upon to maintain the primary 

system’s pressure and inventory. As part 

of the reactor coolant pressure 

boundary, the SG tubes are unique in 

that they are also relied upon as a heat 

transfer surface between the primary 

and secondary systems such that 

residual heat can be removed from the 

primary system. In addition, the SG 

tubes isolate the radioactive fission 

products in the primary coolant from 

the secondary system. In summary, the 

safety function of a SG is maintained by 

ensuring the integrity of its tubes. 

Steam generator tube integrity is a 

function of the design, environment, 

and the physical condition of the tube. 

The proposed change does not affect 

tube design or operating environment. 

The proposed change is expected to 

result in an improvement in the tube 

integrity by implementing the Steam 

Generator Program to manage SG tube 

inspection, assessment, repair, and 

plugging. The requirements established 

by the Steam Generator Program are 

consistent with those in the applicable 

design codes and standards and are an 

improvement over the requirements in 

the current technical specifications. 

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: N.S. Reynolds, 

Esquire, Winston & Strawn 1700 K 

Street, NW., Washington, DC 20006–

3817. 

NRC Section Chief: David Terao. 

Florida Power and Light Company, 

Docket Nos. 50–250 and 50–251, Turkey 

Point Plant, Units 3 and 4, Miami-Dade 

County, Florida 

Date of amendment request: July 21, 

2005. 

Description of amendment request: 

The requested change will delete 

Technical Specification (TS) 6.9.1.2 

related to Occupational Radiation 

Exposure Reports and TS 6.9.1.5, 

‘‘Monthly Operating Reports.’’

Basis for proposed no significant 

hazards consideration determination: 

The NRC staff issued a notice of 

availability of a model no significant 

hazards consideration (NSHC) 

determination for referencing in license 

amendment applications in the Federal 

Register on June 23, 2004 (69 FR 35067). 

The licensee affirmed the applicability 

of the model NSHC determination in its 

application dated July 21, 2005. 

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 an accident 

previously evaluated? 

Response: No. The proposed change 

eliminates the Technical Specifications 

(TSs) reporting requirements to provide 

a monthly operating letter report of 

shutdown experience and operating 

statistics if the equivalent data is 

submitted using an industry electronic 

database. It also eliminates the TS 

reporting requirement for an annual 

occupational radiation exposure report, 

which provides information beyond that 

specified in NRC regulations. The 

proposed change involves no changes to 

plant systems or accident analyses. As 

such, the change is administrative in 

nature and does not affect initiators of 

analyzed events or assumed mitigation 

of accidents or transients. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed change create 

the possibility of a new or different kind 

of accident from any accident 

previously evaluated? 

Response: No. The proposed change 

does not involve a physical alteration of 

the plant, add any new equipment, or 

require any existing equipment to be 

operated in a manner different from the 

present design. Therefore, the proposed 

change does not create the possibility of 

a new or different kind of accident from 

any accident previously evaluated. 

Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 61661 

3. Does the proposed change involve 

a significant reduction in a margin of 

safety? 

Response: No. This is an 

administrative change to reporting 

requirements of plant operating 

information and occupational radiation 

exposure data, and has no effect on 

plant equipment, operating practices or 

safety analyses assumptions. For these 

reasons, the proposed change does not 

involve a significant reduction in the 

margin of safety. 

Based upon the reasoning presented 

above, the requested change does not 

involve significance hazards 

consideration. 

Attorney for licensee: M.S. Ross, 

Attorney, Florida Power & Light, P.O. 

Box 14000, Juno Beach, Florida 33408–

0420. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Florida Power and Light Company, et 

al., Docket Nos. 50–335 and 50–389, St. 

Lucie Plant, Unit Nos. 1 and 2, St. Lucie 

County, Florida 

Date of amendment request: 

September 1, 2005. 

Description of amendment request: 

The requested change will delete 

Technical Specification (TS) 6.9.1.2 

related to Occupational Radiation 

Exposure Reports and TS 6.9.1.6, 

‘‘Monthly Operating Reports.’’

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 eliminates 

the Technical Specifications (TSs) 

reporting requirements to provide a 

monthly operating report of shutdown 

experience and operating statistics if the 

equivalent data is submitted using an 

industry electronic database. It also 

eliminates the TS reporting requirement 

for an annual occupational radiation 

exposure report, which provides 

information beyond that specified in 

NRC regulations. The proposed change 

involves no changes to plant systems or 

accident analyses. As such, the change 

is administrative in nature and does not 

affect initiators of analyzed events or 

assumed mitigation of accidents or 

transients. Therefore, the proposed 

change does not involve a significant 

increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed change create 

the possibility of a new or different kind 

of accident from any accident 

previously evaluated? 

No. The proposed change does not 

involve a physical alteration of the 

plant, add any new equipment, or 

require any existing equipment to be 

operated in a manner different from the 

present design. 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. This is an administrative change 

to reporting requirements of plant 

operating information and occupational 

radiation exposure data, and has no 

effect on plant equipment, operating 

practices or safety analyses 

assumptions. For these reasons, the 

proposed change does not involve a 

significant reduction in the margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: M.S. Ross, 

Attorney, Florida Power & Light, P.O. 

Box 14000, Juno Beach, Florida 33408–

0420. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: July 21, 

2005. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) testing 

frequency for the surveillance 

requirement (SR) in TS 3.1.4, ‘‘Control 

Rod Scram Times.’’ Specifically, the 

proposed change would revise the 

frequency for 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.’’

The NRC staff issued a notice of 

availability of a model no significant 

hazards consideration (NSHC) 

determination for referencing in 

licensing amendment applications in 

the Federal Register on August 23, 2004 

(69 FR 51864). The licensee affirmed the 

applicability of the model NSHC 

determination in its application dated 

July 21, 2005. 

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. 

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 R. 

McClure, Nebraska Public Power 

District, Post Office Box 499, Columbus, 

NE 68602–0499. 

61662 Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 

NRC Section Chief: David Terao. 

Southern Nuclear Operating Company, 

Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of Georgia, 

City of Dalton, Georgia, Docket Nos. 50–

321 and 50–366, Edwin I. Hatch Nuclear 

Plant, Units 1 and 2, Appling County, 

Georgia 

Date of amendment request: 

November 12, 2004, as supplemented by 

letters dated September 2 and 

September 16, 2005. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specifications 3.1.7, 

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

for Hatch, Units 1 and 2. The proposed 

amendments would update Figures 

3.1.7–1 and 3.1.7–2 for Units 1 and 2 TS 

to reflect the increased concentration of 

Boron-10 in the solution. Conforming 

revisions to Bases B 3.1.7, ‘‘Standby 

Liquid Control (SLC) System’’ are also 

included. 

The proposed amendment was 

previously noticed on February 1, 2005 

(70 FR 5249). 

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? 

This is a proposed change to Figures 

3.1.7–1 and 3.1.7–2 of the Units 1 and 

2 TS [Technical Specifications]. Figure 

3.1.7–1 is a plot of the weight percent 

of Sodium Pentaborate solution in the 

Standby Liquid Control (SLC) Tank, as 

a function of the gross volume of 

solution in the tank. Figure 3.1.7–2 is a 

plot of the Sodium Pentaborate 

temperature versus concentration 

requirements. 

Figure 3.1.7–1 is proposed to be 

changed in order to accommodate an 

injection of Sodium Pentaborate 

solution into the reactor, following an 

ATWS [anticipated transient without 

scram] event, such that the 

concentration of Boron-10 atoms in the 

reactor will be 800 ppm natural Boron 

equivalent. This is necessary to 

accommodate increased cycle energy 

requirements for the Hatch Units 1 and 

2 cores. Both Figures 3.1.7–1 and 3.1.7–

2 are changed to reflect that the 

boundary between Region A and B is 

changing from 6.9% to 7.0%. The 

proposed change to the Figures will not 

increase the probability of an ATWS 

event because the curves have nothing 

to do with the prevention of an ATWS 

event. The new requirements will insure 

that, in the future, the core will have 

adequate shutdown margin to mitigate 

the consequences of an ATWS event. 

The minimum concentration of 

Sodium Pentaborate which also 

represents the boundary between Region 

A and Region B, is changing from 6.9% 

to 7.0%. This increase in the 

concentration ensures a conservative 

margin to the ATWS equivalency 

determination required by 10 CFR 

50.62. 

Also, no systems or components 

designed to ensure the safe shutdown of 

the reactor are being physically changed 

as a result of this proposed TS change. 

In fact, no safety related systems or 

components designed for the prevention 

of previously evaluated events are being 

altered by the amendment. 

2. The proposed change does not 

create the possibility of a new or 

different kind of accident from any 

previously evaluated. 

This proposed TS revision results in 

a change to SLC TS Figures 3.1.7–1 and 

3.1.7–2 requirements. However, these 

changes do not result in physical 

changes to the SLC system. SLC pump 

operation, maintenance and testing 

remain the same. Accordingly, no 

changes to the operation, maintenance 

or surveillance procedures will result 

from this TS revision request. Therefore, 

no new modes of operation are 

introduced by this TS change. 

Since no new modes of operation are 

introduced, the proposed change does 

not create the possibility of a new or 

different type event from any previously 

evaluated. 

3. The proposed change does not 

involve a significant reduction in the 

margin of safety. 

This proposed TS change is being 

made to increase the boron 

concentration requirements of the 

sodium pentaborate solution injected 

into the reactor vessel following an 

Anticipated Transient Without Scram 

(ATWS) event. The change is necessary 

due to new fuel designs and higher 

energy requirements for fuel cycles. 

Therefore, the change is being made to 

insure that shutdown requirements can 

be met for the ATWS event. This will 

insure the margin of safety with respect 

to ATWS will continue to be met. 

The increase in the minimum 

concentration from 6.9% to 7.0% 

ensures a conservative margin with 

respect to the ATWS equivalency 

determination. Consequently, this 

proposed TS change will not result in a 

decrease in the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ernest L. Blake, 

Jr., Esquire, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037. 

NRC Section Chief: Evangelos C. 

Marinos. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant, Units 1 and 2, Hamilton 

County, Tennessee 

Date of amendment request: 

September 1, 2005 (TS–05–04). 

Description of amendment request: 

The proposed amendment would revise 

the reactor protection system turbine 

trip allowable value for low trip system 

pressure from greater than or equal to 43 

pounds per square inch gauge (psig) to 

39.5 psig. This change would allow the 

instrumentation that performs this trip 

function to be tested and verified to be 

operable within the capabilities of the 

pressure switches. 

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 

revises the allowable value for reactor 

trip as a result of a turbine trip on low 

trip system pressure. This change will 

not alter any plant components, 

systems, or processes and will only 

provide a more appropriate value to 

assess operability of the associated 

pressure switches. Since the plant 

features and operating practices are not 

altered, the possibility of an accident is 

not affected. This reactor trip is not 

directly credited in SQN’s accident 

analysis and is maintained as an 

anticipatory trip to enhance the overall 

reliability of the reactor trip system. As 

such, there is not a specific safety limit 

associated with this function and the 

generation of a reactor trip based on low 

trip system pressure is above the 

required actuations to ensure acceptable 

mitigation of accidents. As the proposed 

change will continue to provide an 

acceptable anticipatory trip signal, the 

offsite dose potential is not affected by 

this change. Therefore, the proposed 

change does not involve a significant 

increase in the probability or 

Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 61663 

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. As described above, 

this change will not alter any plant 

equipment or operating practices that 

have the ability to create a new potential 

for accident generation. The proposed 

change revises the operability limits for 

a function that generates a trip signal 

when appropriate conditions exist to 

require accident mitigation response. 

This type of function does not have the 

ability to create an accident as its 

purpose and function is to mitigate 

events. 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 

will revise an allowable value for a 

reactor trip initiator that results from a 

turbine trip condition. This change will 

not alter the setpoint, and the 

calibration of the associated pressure 

switches will continue to be set at the 

current values. The allowable value 

change is in response to accuracy 

aspects of the instrumentation and does 

not alter the ability of this trip function 

to operate when and as needed to 

mitigate accident conditions. 

Therefore, the proposed change does 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: August 

26, 2005. 

Description of amendment request: 

The amendment would authorize 

changes to the Updated Safety Analysis 

Report (USAR) for Wolf Creek 

Generating Station (WCGS) that would 

revise the methodology for the reactor 

coolant system (RCS) leak detection 

instrumentation. This revision would 

clarify the requirements of the 

containment atmosphere gaseous 

radioactivity monitor with regard to the 

RCS leak detection capability and 

would justify that the monitor can be 

considered operable in compliance with 

Limiting Condition for Operation 3.4.15, 

in Technical Specification (TS) 3.4.15, 

‘‘RCS Leakage Detection 

Instrumentation,’’ during all applicable 

Modes. There are no proposed changes 

to the WCGS 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. The proposed change does not 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated. 

The proposed change has been 

evaluated and determined to not 

increase the probability or consequences 

of an accident previously evaluated. The 

proposed change does not make 

hardware changes and does not alter the 

configuration of any plant system, 

structure, or component (SSC). The 

proposed change only clarifies the 

design and OPERABILITY requirements 

for the containment atmosphere gaseous 

radioactivity monitors and identifies the 

capabilities of the monitors at low RCS 

[radio]activity levels. The containment 

atmosphere gaseous radioactivity 

monitors are not initiators of any 

accident; therefore, the probability of 

occurrence of an accident is not 

increased. The USAR and TSs will 

continue to require diverse means of 

[RCS] leakage detection equipment, thus 

ensuring that leakage due to cracks [in 

the RCS] would continue to be 

identified prior to propagating to the 

point of a[n] [RCS] pipe break. 

Therefore, the consequences of an 

accident [previously evaluated] are not 

increased. 

2. The proposed change does not 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated. 

The proposed change does involve the 

use or installation of new equipment 

and the currently installed equipment 

will not be operated in a new or 

different manner. No new or different 

system interactions are created and no 

new processes are introduced. The 

proposed changes will not introduce 

any new failure mechanisms, 

malfunctions, or accident initiators not 

already considered in the design and 

licensing basis [for WCGS]. The 

proposed change does not affect any 

SSC associated with an accident 

initiator. Based on this evaluation, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. The proposed change does not 

involve a significant reduction in a 

margin of safety. 

The proposed change does not alter 

any RCS leakage detection components. 

The proposed change only clarifies the 

design and operability requirements for 

the containment atmosphere gaseous 

radioactivity monitor and identifies the 

capabilities of the containment 

atmosphere gaseous radioactivity 

monitors at low RCS [radio]activity 

levels. This change is required since the 

level of radioactivity in the WCGS 

reactor coolant has become much lower 

than what was assumed in the USAR 

and the gaseous channel [(monitor)] can 

no longer promptly detect a small RCS 

leak under all operating conditions. The 

proposed amendment continues to 

require diverse means of [RCS] leakage 

detection equipment with [the] 

capability to promptly detect RCS 

leakage. Although not required by [the] 

TS[s], additional diverse means of 

leakage detection capability are 

available as described in the USAR 

Section 5.2.5. Early detection of [RCS] 

leakage, as the potential indicator of a 

crack(s) in the RCS pressure boundary, 

will thus continue to be in place so that 

such a condition is known and 

appropriate actions [are] taken well 

before any such crack would propagate 

to a more severe condition. Based on 

this evaluation, 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: Jay Silberg, Esq., 

Shaw, Pittman, Potts and Trowbridge, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Section Chief: Daniel S. Collins, 

Acting. 

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 

61664 Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

Systems (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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. 50–289, Three Mile Island Nuclear 

Station, Unit 1 (TMI–1), Dauphin 

County, Pennsylvania 

Date of application for amendment: 

October 20, 2004, as supplemented by 

letters dated June 30, July 29, August 17, 

and September 19, 2005. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications to (1) eliminate the 

existing requirement in Section 3.8.6 

regarding maintaining the containment 

equipment hatch cover in place with a 

minimum of four bolts during fuel 

loading and refueling operations, and 

(2) revise or introduce commitments to 

the Technical Specifications Bases in 

support of the change in Section 3.8.6. 

Date of issuance: October 13, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 257. 

Facility Operating License No. DPR–

50. Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 7, 2004 (69 FR 

70714) The supplements dated June 30, 

July 29, August 17, and September 19, 

2005, 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 

October 13, 2005. 

No significant hazards consideration 

comments received: No. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units Nos. 1, 2, and 

3, Maricopa County, Arizona 

Date of application for amendments: 

May 28, 2003, as supplemented by 

letters dated January 22 and June 23, 

2004, and February 2 and September 27, 

2005. 

Brief description of amendments: The 

amendments revise several surveillance 

requirements (SRs) in Technical 

Specification (TS) 3.8.1 on alternating 

current power sources and SR 3.8.4.6 for 

direct current power sources for plant 

operation. The revised SRs have notes 

deleted or modified to adopt in part the 

staff-approved TS Task Force 283, 

Revision 3, to allow these SRs to be 

performed, or partially performed, in 

reactor modes that previously were not 

allowed by the TSs. 

Date of issuance: September 29, 2005. 

Effective date: September 29, 2005, 

and shall be implemented within 90 

days of the date of issuance including 

the incorporation of the changes to the 

TS Bases for TS 3.8.1 and SR 3.8.4.6 as 

described in the licensee’s letters dated 

May 28, 2003, January 22 and June 23, 

2004, and February 2 and September 27, 

2005. 

Amendment Nos.: Unit 1—156, Unit 

2—156, Unit 3—156. 

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The 

amendments revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 8, 2003 (68 FR 40709). 

The supplemental letters dated 

January 22, June 23, 2004, and February 

2 and September 27, 2005, do not 

expand the scope of the application as 

originally noticed and do 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 September 29, 

2005. 

No significant hazards consideration 

comments received: No. 

Dominion Energy Kewaunee, Inc. Docket 

No. 50–305, Kewaunee Power Station, 

Kewaunee County, Wisconsin 

Date of application for amendment: 

February 3, 2005. 

Brief description of amendment: The 

amendment modified Technical 

Specification (TS) 6.16.b.1, ‘‘Radioactive 

Effluent Controls Program,’’ and TS 

6.18, ‘‘Off-site Dose Calculation Manual 

(ODCM),’’ to be consistent with Title 10 

of the Code of Federal Regulations (10 

CFR) part 20 and NUREG–1431, 

‘‘Standard Technical Specifications 

Westinghouse Plants.’’

Date of issuance: October 4, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 186. 

Facility Operating License No. DPR–

43: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: March 29, 2005 (70 FR 

15944). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 4, 2005. 

No significant hazards consideration 

comments received: No. 

Duke Energy Corporation, 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: 

November 25, 2002, as supplemented by 

letters dated November 13 and 

December 16, 2003, September 22, 2004, 

April 6, June 14, July 8, August 17, and 

September 8 and September 19, 2005. 

Brief description of amendments: The 

amendments include a full-scope 

implementation of an alternative source 

term for evaluating the consequences of 

design basis accidents at Catawba 

Nuclear Station. The amendments also 

revised the Technical Specifications for 

the Ventilation Filter Testing Program, 

Annulus Ventilation System, Auxiliary 

Building Filtered Ventilation Exhaust 

Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 61665 

System, Fuel Handling Ventilation 

Exhaust System, and Control Room Area 

Ventilation System, and containment 

penetrations. 

Date of issuance: September 30, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 227 and 222. 

Renewed Facility Operating License 

Nos. NPF–35 and NPF–52: Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: April 15, 2003 (68 FR 18272). 

This application was renoticed on May 

24, 2005 (70 FR 29789). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 30, 

2005. 

No significant hazards consideration 

comments received: No. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of application for amendment: 

April 19, 2005. 

Brief description of amendment: The 

amendment revised technical 

specifications (TSs) testing frequency 

for the surveillance requirement (SR) in 

TS 3.1.4, ‘‘Control Rod Scram Times.’’

Specifically, the change revised the 

frequency for 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.’’

Date of issuance: September 29, 2005. 

Effective date: September 29, 2005, 

and shall be implemented within 60 

days from the date of issuance. 

Amendment No.: 194. 

Facility Operating License No. NPF–

21: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: June 7, 2005 (70 FR 33212). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 29, 

2005. 

No significant hazards consideration 

comments received: No. 

Entergy Gulf States, Inc., and Entergy 

Operations, Inc., Docket No. 50–458, 

River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: 

December 17, 2004, as supplemented by 

letters dated June 29, and August 12, 

2005. 

Brief description of amendment: The 

amendment revises the Technical 

Specification (TS) requirements for 

direct current (DC) sources. The current 

TS only includes ACTION Statements 

for an inoperable DC Power subsystems. 

The change adds a new ACTION 

Statement to TS 3.8.4, ‘‘DC Sources—

Operating,’’ to specifically address an 

inoperable battery charger. 

Date of issuance: October 7, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment No.: 148. 

Facility Operating License No. NPF–

47: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: January 4, 2005 (70 FR 401). 

The supplements dated June 29, and 

August 12, 2005, 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 October 7, 2005. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2, 

Pope County, Arkansas 

Date of application for amendment: 

December 20, 2004, as supplemented by 

letter dated April 26, 2005. 

Brief description of amendment: The 

amendment changed the existing 

containment structures and tendon 

inservice inspection requirements to be 

consistent with NUREG–1432, Revision 

3, and the American Society of 

Mechanical Engineers Boiler and 

Pressure Vessel Code, Section XI. 

Specifically, the amendment modified 

the Surveillance Requirement of 

Technical Specification (TS) 3.6.1.5, 

added a new Surveillance Program to 

TS 6.5.6 and a report to TS 6.5.7, and 

made two administrative changes to the 

TSs. 

Date of issuance: September 29, 2005. 

Effective date: As of the date of 

issuance to be implemented within 90 

days from the date of issuance. 

Amendment No.: 262. 

Renewed Facility Operating License 

No. NPF–6: Amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: March 29, 2005 (70 FR 

15943). 

The supplement dated April 26, 2005, 

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 September 29, 

2005. 

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: 

September 21, 2004, as supplemented 

by letters dated March 18, April 7, May 

6, August 10, and September 19, 2005. 

Brief description of amendments: The 

amendments extended the outage times 

from 72 hours to 14 days for an 

inoperable emergency diesel generator. 

It also changed formats of the affected 

technical specification pages to improve 

their appearance but not alter any 

requirements. 

Date of issuance: September 30, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days. 

Amendment Nos.: 291, 273. 

Facility Operating License Nos. DPR–

58 and DPR–74: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: October 26, 2004 (69 FR 

62476). The supplements provided 

clarifying information that did not 

change the scope of the proposed 

amendment as described in the original 

notice of proposed action published in 

the Federal Register, and did not 

change the initial proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 30, 

2005. 

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: 

June 30, 2004, as supplemented by 

letters dated September 16, 2004, 

November 5, 2004, March 3, 2005, July 

1, 2005, and September 27, 2005. 

Brief description of amendment: The 

amendment changed the TSs to support 

an increase in the length of the fuel 

cycle from 18 to 24 months at 

Monticello. In addition, the proposed 

amendment requested changes in 

calibration times of various instruments. 

These changes will be evaluated in a 

separate license amendment. 

61666 Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 

Date of issuance: September 30, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 143. 

Facility Operating License No. DPR–

22: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: January 18, 2005 (70 FR 

2892). The supplements dated 

September 16, 2004, November 5, 2004, 

March 3, 2005, July 1, 2005, and 

September 27, 2005, 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 on 

January 18, 2005 (70 FR 2892). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 30, 

2005. 

No significant hazards consideration 

comments received: No. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request: 

November 23, 2004, as supplemented by 

letter dated July 8, 2005. 

Brief description of amendment: The 

amendment (1) revised the descriptive 

wording of Technical Specifications 

(TSs) Table 1–1, ‘‘RPS [reactor 

protection system] Limiting Safety 

System Settings,’’ for the reactor trip 

setpoint for low steam generator water 

level to relocate unnecessary detail, and 

(2) converted TSs Section 4.0, ‘‘Design 

Features,’’ to the format and content of 

NUREG–1432, Revision 3, ‘‘Standard 

Technical Specifications for 

Combustion Engineering Plants.’’

Date of issuance: October 3, 2005. 

Effective date: October 3, 2005, and 

shall be implemented within 60 days 

from the date of issuance. 

Amendment No.: 236. 

Renewed Facility Operating License 

No. DPR–40: The amendment revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: May 24, 2005 (70 FR 29798). 

The July 8, 2005, supplemental letter 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

safety evaluation dated October 3, 2005. 

No significant hazards consideration 

comments received: No. 

PPL Susquehanna, LLC, Docket Nos. 50–

387 and 50–388, Susquehanna Steam 

Electric Station, Units 1 and 2 (SSES 1 

and 2), Luzerne County, Pennsylvania 

Date of application for amendments: 

September 8, 2004, as supplemented by 

letters dated July 8 and September 28, 

2005. 

Brief description of amendments: The 

amendments changed SSES 1 and 2 

Technical Specifications 3.6.4.1, 

‘‘Secondary Containment,’’ and 3.6.4.3, 

‘‘Standby Gas Treatment System 

(SGTS),’’ to extend, on a one-time basis, 

the allowable completion time for 

required actions for secondary 

containment inoperable and two SGTS 

subsystems inoperable, in mode 1, 2, or 

3, from 4 hours to 48 hours. 

Date of issuance: October 6, 2005. 

Effective date: October 6, 2005. 

Amendment Nos.: 226 and 203. 

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: March 1, 2005 (70 FR 9994). 

The supplements dated July 8 and 

September 28, 2005, contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination or expand 

the scope of the initial Federal Register 

notice. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 6, 2005. 

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: 

April 15, 2004, as supplemented by 

letters dated August 11, 2004, and 

August 11, 2005. The August 11, 2005, 

supplement withdrew a portion of the 

original application from consideration. 

Brief description of amendments: The 

amendments revised Technical 

Specification (TS) 3.6.2.3 Action B, for 

both units, to correct a non-conservative 

action statement. 

Date of issuance: September 30, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 266 and 248. 

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: October 12, 2004 (69 FR 

60684). The licensee’s supplement 

dated August 11, 2005, withdrew a 

portion of the original application from 

consideration and did not increase the 

scope of the initial proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 30, 

2005. 

No significant hazards consideration 

comments received: No. 

Southern California Edison Company, et 

al., Docket Nos. 50–361 and 50–362, 

San Onofre Nuclear Generating Station, 

Units 2 and 3, San Diego County, 

California 

Date of application for amendments: 

March 24, 2005. 

Brief description of amendments: The 

proposed changes revised various 

Technical Specifications (TSs) related to 

cycle-specific values and the shutdown 

margin, and are consistent with the 

following Nuclear Regulatory 

Commission approved Technical 

Specification Task Force (TSTF) 

Standard TS Change Travelers: TSTF–

9–A, Revision 1, ‘‘Relocate Value for 

Shutdown Margin to COLR;’’ TSTF–67–

A, Revision 0, ‘‘Correction of Shutdown 

Margin Definition;’’ TSTF–142–A, 

Revision 0, ‘‘Increase the Completion 

Time When the Core Reactivity Balance 

is Not Within Limit;’’ and TSTF–150–A, 

Revision 0, ‘‘Replace DNBR Power 

Decrease Number with Reference to the 

COLR.’’

Date of issuance: October 3, 2005. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: 200/191. 

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: May 10, 2005 (70 FR 24656). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 3, 2005. 

No significant hazards consideration 

comments received: No. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: July 4, 

2005. 

Brief description of amendments: The 

amendments change Technical 

Specification 4.0.5 to add a reference to 

the NRC-approved exemption of 

selected pumps, valves, and other 

components from special treatment 

requirements. As an editorial change, 

references to Title 10, Code of Federal 

Regulations (10 CFR) part 50, section 

Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 61667 

50.55a(f) and 10 CFR part 50, section 

50.55a(f)(6)(I) is added to the paragraph 

for inservice testing, similar to the 

existing references for inservice 

inspection. In addition, ‘‘inservice 

testing’’ and ‘‘inservice inspection’’ are 

reordered for consistency with the 

sequence of the regulations in 10 CFR 

50.55a. 

Date of issuance: October 4, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days of issuance. 

Amendment Nos.: Unit 1—173; Unit 

2—161. 

Facility Operating License Nos. NPF–

76 and NPF–80: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: August 2, 2005 (70 FR 44403). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 4, 2005. 

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: 

September 9, 2004. 

Brief description of amendments: The 

Amendments revised the Technical 

Specification 3.6.6.8 to change the 

current interval for surveillance from 

every 10 years to verification that the 

nozzles are unobstructed following a 

maintenance that could have resulted in 

nozzle blockage. 

Date of issuance: September 23, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: 120 and 120. 

Facility Operating License Nos. NPF–

87 and NPF–89: The Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: October 26, 2004 (69 FR 

62478). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 23, 

2005. 

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: 

September 17, 2004, as supplemented 

by letters dated February 11, May 26, 

June 17 (two letters), July 15, July 29, 

August 16, and September 6, 2005. 

Brief description of amendment: The 

amendment supports the installation of 

replacement steam generators (SGs) at 

Callaway during the refueling outage in 

the fall of 2005. The amendment affects 

the following affected TSs: the reactor 

core safety limits (TS 2.1.1), reactor trip 

system and engineered safety feature 

actuation system instrumentation (TSs 

3.3.1 and 3.3.2), reactor coolant system 

(RCS) limits (TS 3.4.1), RCS loops (TSs 

3.4.5, 3.4.6, and 3.4.7), RCS operational 

leakage (TS 3.4.13), SG tube integrity 

(the new TS 3.4.17), main steam safety 

valves (TS 3.7.1), SG tube surveillance 

program (TS 5.5.9), containment 

integrated leakage rate testing program 

(TS 5.5.16), and SG tube inspection 

report (TS 5.6.10). 

Date of issuance: September 29, 2005. 

Effective date: Effective on the date of 

issuance, and shall be implemented 

before entry into Mode 5 during the 

restart from the fall 2005 refueling 

outage when the replacement steam 

generators are installed including (1) 

revising the pressure temperature limits 

report to change the cold overpressure 

mitigation system setpoints to reflect no 

reactor coolant pump operation 

restrictions and (2) incorporating the TS 

Bases changes identified in the 

licensee’s letter of September 6, 2005, 

into the TS Bases. 

Amendment No.: 168. 

Facility Operating License No. NPF–

30: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: November 23, 2004 (69 FR 

68185). The supplemental letters dated 

February 11, May 26, June 17 (two 

letters), July 15, July 29, August 16, and 

September 6, 2005, 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. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 29, 

2005. 

No significant hazards consideration 

comments received: No. 

Notice of Issuance of Amendments to 

Facility Operating Licenses and Final 

Determination of No Significant 

Hazards Consideration and 

Opportunity for a Hearing (Exigent 

Public Announcement or Emergency 

Circumstances) 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application for the 

amendment complies with the 

standards and requirements of the 

Atomic Energy Act of 1954, as amended 

(the Act), and the Commission’s rules 

and regulations. The Commission has 

made appropriate findings as required 

by the Act and the Commission’s rules 

and regulations in 10 CFR Chapter I, 

which are set forth in the license 

amendment. 

Because of exigent or emergency 

circumstances associated with the date 

the amendment was needed, there was 

not time for the Commission to publish, 

for public comment before issuance, its 

usual Notice of Consideration of 

Issuance of Amendment, Proposed No 

Significant Hazards Consideration 

Determination, and Opportunity for a 

Hearing. 

For exigent circumstances, the 

Commission has either issued a Federal 

Register notice providing opportunity 

for public comment or has used local 

media to provide notice to the public in 

the area surrounding a licensee’s facility 

of the licensee’s application and of the 

Commission’s proposed determination 

of no significant hazards consideration. 

The Commission has provided a 

reasonable opportunity for the public to 

comment, using its best efforts to make 

available to the public means of 

communication for the public to 

respond quickly, and in the case of 

telephone comments, the comments 

have been recorded or transcribed as 

appropriate and the licensee has been 

informed of the public comments. 

In circumstances where failure to act 

in a timely way would have resulted, for 

example, in derating or shutdown of a 

nuclear power plant or in prevention of 

either resumption of operation or of 

increase in power output up to the 

plant’s licensed power level, the 

Commission may not have had an 

opportunity to provide for public 

comment on its no significant hazards 

consideration determination. In such 

case, the license amendment has been 

issued without opportunity for 

comment. If there has been some time 

for public comment but less than 30 

days, the Commission may provide an 

opportunity for public comment. If 

comments have been requested, it is so 

stated. In either event, the State has 

been consulted by telephone whenever 

possible. 

Under its regulations, the Commission 

may issue and make an amendment 

immediately effective, notwithstanding 

the pendency before it of a request for 

a hearing from any person, in advance 

of the holding and completion of any 

required hearing, where it has 

determined that no significant hazards 

consideration is involved. 

61668 Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 

1To the extent that the applications contain 

attachments and supporting documents that are not 

publicly available because they are asserted to 

contain safeguards or proprietary information, 

petitioners desiring access to this information 

should contact the applicant or applicant’s counsel 

and discuss the need for a protective order. 

The Commission has applied the 

standards of 10 CFR 50.92 and has made 

a final determination that the 

amendment involves no significant 

hazards consideration. The basis for this 

determination is contained in the 

documents related to this action. 

Accordingly, the amendments have 

been issued and made effective as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the application for 

amendment, (2) the amendment to 

Facility Operating License, and (3) the 

Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment, as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

reading-rm/adams.html. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

The Commission is also offering an 

opportunity for a hearing with respect to 

the issuance of the amendment. Within 

60 days after the date of publication of 

this notice, the licensee may file a 

request for a hearing with respect to 

issuance of the amendment to the 

subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

intervene. Requests for a hearing and a 

petition for leave to intervene shall be 

filed in accordance with the 

Commission’s ‘‘Rules of Practice for 

Domestic Licensing Proceedings’’ in 10 

CFR part 2. Interested persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland, 

and electronically on the Internet at the 

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

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

are problems in accessing the document, 

contact the PDR Reference staff at 1 

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

hearing or petition for leave to intervene 

is filed by the above date, the 

Commission or a presiding officer 

designated by the Commission or by the 

Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner intends to 

rely in proving the contention at the 

hearing. The petitioner must also 

provide references to those specific 

sources and documents of which the 

petitioner is aware and on which the 

petitioner intends to rely to establish 

those facts or expert opinion. The 

petition must include sufficient 

information to show that a genuine 

dispute exists with the applicant on a 

material issue of law or fact.1

Contentions shall be limited to matters 

within the scope of the amendment 

under consideration. The contention 

must be one which, if proven, would 

entitle the petitioner to relief. A 

petitioner/requestor who fails to satisfy 

these requirements with respect to at 

least one contention will not be 

permitted to participate as a party. 

Each contention shall be given a 

separate numeric or alpha designation 

within one of the following groups: 

1. Technical: primarily concerns/ 

issues relating to technical and/or 

health and safety matters discussed or 

referenced in the applications. 

2. Environmental: primarily concerns/ 

issues relating to matters discussed or 

referenced in the environmental 

analysis for the applications. 

3. Miscellaneous: does not fall into 

one of the categories outlined above. 

As specified in 10 CFR 2.309, if two 

or more petitioners/requestors seek to 

co-sponsor a contention, the petitioners/ 

requestors shall jointly designate a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. If a petitioner/requestor 

seeks to adopt the contention of another 

sponsoring petitioner/requestor, the 

petitioner/requestor who seeks to adopt 

the contention must either agree that the 

sponsoring petitioner/requestor shall act 

as the representative with respect to that 

contention, or jointly designate with the 

sponsoring petitioner/requestor a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. Since the Commission has 

made a final determination that the 

amendment involves no significant 

hazards consideration, if a hearing is 

requested, it will not stay the 

effectiveness of the amendment. Any 

hearing held would take place while the 

amendment is in effect. 

A request for a hearing or a petition 

for leave to intervene must be filed by: 

(1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

Federal Register / Vol. 70, No. 205 / Tuesday, October 25, 2005 / Notices 61669 

0001, Attention: Rulemaking and 

Adjudications Staff; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) E-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HearingDocket@nrc.gov; or (4) facsimile 

transmission addressed to the Office of 

the Secretary, U.S. Nuclear Regulatory 

Commission, Washington, DC, 

Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

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

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer or 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(I)–(viii). 

Entergy Gulf States, Inc., and Entergy 

Operations, Inc., Docket No. 50–458, 

River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: August 

31, 2005, as supplemented by letter 

dated September 13, 2005. 

Brief description of amendment: The 

amendment permitted a one-time 

change to Technical Specification Table 

3.3.8.1–1 to provide a one-time 

relaxation of the Loss of Power 

instrumentation requirements. 

Date of issuance: September 15, 2005. 

Effective date: As of the date of 

issuance to be implemented 

immediately. 

Amendment No.: 147. 

Facility Operating License No. NPF–

47: Amendment revised the Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration: Yes. 

The NRC published a public notice of 

the proposed amendment, issued a 

proposed finding of no significant 

hazards consideration, and requested 

that any comments on the proposed no 

significant hazards consideration be 

provided to the NRC staff by the close 

of business on September 9, 2005. The 

notice was published in The St. 

Francisville Democrat (in St. 

Francisville) on September 8, 2005, and 

The Advocate (in Baton Rouge) on 

September 7, 2005. No public comments 

were received. 

The Commission’s related evaluation 

of the amendment, finding of exigent 

circumstances, consultation with the 

State of Louisiana, and final no 

significant hazards consideration 

determination are contained in a Safety 

Evaluation dated September 15, 2005. 

XU Generation Company LP, Docket No. 

50–445, Comanche Peak Steam Electric 

Station, Unit No. 1, Somervell County, 

Texas 

Date of amendment request: April 27, 

2005 as supplemented by letter dated 

July 20, 2005. 

Description of amendment: The 

amendments revise the Technical 

Specifications to add the topical report 

WCAP–13060–P–A to the list of NRC 

approved methodologies to be used at 

Comanche Peak Steam Electric Station, 

Unit 1. 

Date of issuance: October 11, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

immediately. 

Amendment No.: 123. 

Facility Operating License No. NPF–

87: Amendment revised the Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): Yes. 

The notice published on September 

26, 2005 (70 FR 56191) provided an 

opportunity to submit comments on the 

Commission’s proposed NSHC 

determination. No comments have been 

received. The notice also provided an 

opportunity to request a hearing within 

60 days from the date of publication, but 

indicated that if the Commission makes 

a final NSHC determination, any such 

hearing would take place after issuance 

of the amendment. 

The Commission’s related evaluation 

of the amendment, finding of exigent 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated October 11, 

2005. 

Attorney for licensee: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Section Chief: David Terao. 

Dated at Rockville, Maryland, this 17th day 

of October, 2005. 

For the Nuclear Regulatory Commission. 

Ledyard B. Marsh, 

Director, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 05–21180 Filed 10–24–05; 8:45 am] 

BILLING CODE 7590–01–P 

SECURITIES AND EXCHANGE 

COMMISSION 

Sunshine Act Meeting 

Notice is hereby given, pursuant to 

the provisions of the Government in the 

Sunshine Act, Public Law 94–409, that 

the Securities and Exchange 

Commission will hold the following 

meeting during the week of October 24, 

2005: 

A Closed Meeting will be held on 

Thursday, October 27, 2005 at 2 p.m. 

Commissioners, Counsel to the 

Commissioners, the Secretary to the 

Commission, and recording secretaries 

will attend the Closed Meeting. Certain 

staff members who have an interest in 

the matters may also be present. 

The General Counsel of the 

Commission, or his designee, has 

certified that, in his opinion, one or 

more of the exemptions set forth in 5 

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

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

(10) permit consideration of the 

scheduled matters at the Closed 

Meeting. 

Commissioner Atkins, as duty officer, 

voted to consider the items listed for the 

closed meeting in closed session and 

that no earlier notice thereof was 

possible. 

The subject matters of the Closed 

Meeting scheduled for Thursday, 

October 27, 2005 will be: 

Formal orders of private investigations; 

Institution and settlement of injunctive 

actions; 

Institution and settlement of 

administrative proceedings of an 

enforcement nature; and 

Opinions. 

At times, changes in Commission 

priorities require alterations in the 

scheduling of meeting items. 

For further information and to 

ascertain what, if any, matters have been 

added, deleted or postponed, please 

contact: 

The Office of the Secretary at (202) 

551–5400. 

Dated: October 20, 2005. 

Jonathan G. Katz, 

Secretary. 

[FR Doc. 05–21355 Filed 10–21–05; 11:26 

am] 

BILLING CODE 8010–01–P