70712 Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 

loaded onto an Illyushin IL–76 for the 

return flight to Punta Arenas, Chile. The 

Illyushin is operated by Antarctic 

Logistics and Expeditions (ALE). 

Refueling operations will take place at 

pre-existing fuel caches. 

Application for the permit is made by: 

Steve Brooks, Pole-To-Pole, 1202 Pierce 

100 Street, Clearwater, Florida, 96161. 

Location: Patriot Hills, and South 

Geographic Pole. 

Dates: January 1, 2005 to January 31, 

2005.

Nadene G. Kennedy, 

Permit Officer.

[FR Doc. 04–26826 Filed 12–6–04; 8:45 am] 

BILLING CODE 7555–01–M

NATIONAL SCIENCE FOUNDATION

Notice of Permits Issued Under the 

Antarctic Conservation Act of 1978

AGENCY: National Science Foundation.

ACTION: Notice of permits issued under 

the Antarctic Conservation of 1978, 

Public Law 95–541. 

SUMMARY: The National Science 

Foundation (NSF) is required to publish 

notice of permits issued under the 

Antarctic Conservation Act of 1978. 

This is required notice.

FOR FURTHER INFORMATION CONTACT:

Nadene G. Kennedy, Permit Office, 

Office of Polar Programs, Rm. 755, 

National Science Foundation, 4201 

Wilson Boulevard, Arlington, VA 22230.

SUPPLEMENTARY INFORMATION: On 

September 20, 2004, the National 

Science Foundation published a notice 

in the Federal Register of a Waste 

Management permit application 

received. A Waste Management permit 

was issued on November 30, 2004 to the 

following applicant: Ralph Fedor, Peter 

1st Expedition, Permit No.: 2005 WM–

003.

Nadene G. Kennedy, 

Permit Officer.

[FR Doc. 04–26827 Filed 12–6–04; 8:45 am] 

BILLING CODE 7555–01–M

NUCLEAR REGULATORY 

COMMISSION 

Sunshine Act Meeting

Agency Holding the Meeting: Nuclear 

Regulatory Commission.

Date: Weeks of December 6, 13, 20, 27, 

January 3, 10, 2004.

Place: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland.

Status: Public and Closed.

Matters To Be Considered: Week of 

December 6, 2004

Tuesday, December 7, 2004

9:30 a.m. Briefing on Equal 

Employment Opportunity (EEO) 

Program (Public Meeting) (Contact: 

Corenthis Kelley, (301) 415–7380). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov.

Wednesday, December 8, 2004

12:55 p.m. Affirmation Session 

(Public Meeting) (Tentative) 

a. Motion to Quash OI Subpoena 

(Tentative) 

b. Duke Energy Corp. (Catawba 

Nuclear Station, Units 1 and 2); 

Intervenor’s Motion for 

Reconsideration of CLI–04–29 

(Tentative) 

c. SECY–04–0180—Hydro Resources, 

Inc. (Rio Rancho, New Mexico) 

Review of LBP–04–3 (Financial 

Assurance) (Tentative) 

d. SECY–04–0190—Final Rule: 

Security Requirements for Portable 

Gauges Containing Byproduct 

Material (RIN 3150–AH06) 

(Tentative) 

e. SECY–04–0208—Louisiana Energy 

Services, L.P. (National Enrichment 

Facility) (Tentative) 

f. SECY–04–0212—Dominion Nuclear 

Connecticut, Inc., (Millstone 

Nuclear Power Station, Units 2 and 

3), Docket Nos. 50–336–LR & 50–

423–LR; LBP–04–15, 60 NRC 81, 

LBP–04–22 (Tentative) 

1 p.m. Briefing on Status of Davis 

Besse Lessons Learned Task Force 

Recommendations (Public Meeting) 

(Contact: John Jolicoeur, (301) 415–

1724) 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov.

Thursday, December 9, 2004

2 p.m. Briefing on Reactor Safety and 

Licensing Activities (Public 

Meeting) (Contact: Steve Koenick, 

301–415–1239). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov.

Week of December 13, 2004—Tentative 

Tuesday, December 14, 2004

1 p.m. Briefing on Emergency 

Preparedness Program Initiatives 

(Public Meeting) (Contact: Nader 

Mamish, (301) 415–1086). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov.

Week of December 20, 2004—Tentative 

There are no meetings scheduled for 

the Week of December 20, 2004. 

Week of December 27, 2004—Tentative 

There are no meetings scheduled for 

the Week of December 27, 2004. 

Week of January 3, 2005—Tentative 

There are no meetings scheduled for 

the Week of January 3, 2005. 

Week of January 10, 2005—Tentative 

There are no meetings scheduled for 

the Week of January 10, 2005. 

*The schedule for Commission 

meetings is subject to change on short 

notice. To verify the status of meetings 

call (recording)—(301) 415–1292. 

Contact person for more information: 

Dave Gamberoni, (301) 415–1651.

* * * * *

The NRC Commission Meeting 

Schedule can be found on the Internet 

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

policy-making/schedule.html.

* * * * *

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g.,

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

August Spector, at (301) 415–7080, 

TDD: (301) 415–2100, or by e-mail at 

aks@nrc.gov. Determinations on 

requests for reasonable accommodation 

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

* * * * *

This notice is distributed by mail to 

several hundred subscribers; if you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301–415–1969). 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to dkw@nrc.gov.

Dave Gamberoni, 

Office of the Secretary.

[FR Doc. 04–26899 Filed 12–3–04; 9:27 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 

Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 70713

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 November 5, 

2004, through November 24, 2004. The 

last biweekly notice was published on 

November 23, 2004 (69 FRN 68180). 

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. (Note:

Public access to ADAMS has been 

temporarily suspended so that security 

reviews of publicly available documents 

may be performed and potentially 

sensitive information removed. Please 

check the NRC Web site for updates on 

the resumption of ADAMS access.) 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/. (Note:

Public access to ADAMS has been 

temporarily suspended so that security 

reviews of publicly available documents 

may be performed and potentially 

sensitive information removed. Please 

check the NRC Web site for updates on 

the resumption of ADAMS access.) If a 

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/

requestor seeks to have litigated at the 

proceeding.

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

70714 Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 

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 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. (Note: Public 

access to ADAMS has been temporarily 

suspended so that security reviews of 

publicly available documents may be 

performed and potentially sensitive 

information removed. Please check the 

NRC Web site for updates on the 

resumption of ADAMS access.) If you 

do not have access to ADAMS or if there 

are problems in accessing the 

documents located in ADAMS, contact 

the NRC 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 amendment request: October 

20, 2004. 

Description of amendment request:

The proposed amendment would revise 

the containment hatch closure 

requirement in the Technical 

Specifications (TSs) during fuel 

handling and refueling operations. 

Specifically, the requirement of TS 3.8.6 

that the containment equipment hatch 

remain closed with a minimum of 4 

bolts securing it in place is replaced 

with the requirement that the 

equipment hatch be capable of being 

closed during fuel handling and 

refueling operations. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change is only related to a 

postulated fuel handling accident inside the 

Reactor Building occurring during fuel 

loading and refueling activities. The 

proposed change does not increase the 

probability of a fuel handling accident in that 

the proposed change deals with the results of 

such an accident, not the cause of such an 

accident. The proposed change does not 

increase the consequences of an accident 

previously evaluated in that the TMI Unit 1 

Alternative Source Term has been previously 

reviewed and approved by the NRC [Nuclear 

Regulatory Commission], and this proposed 

change is consistent with the assumptions of 

[that] previous analysis. The Alternative 

Source Term analysis for the Fuel Handling 

Accident [i]nside the Reactor Building takes 

no credit for the closure of the containment 

equipment hatch opening or for a filtered 

release. Previous analyses of external events 

were reviewed and the proposed [change 

does] not affect the conclusions of these 

analyses. Therefore the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not affect nor 

create a different [kind] of fuel handling 

accident. The proposed change is consistent 

with the existing licensing basis accident 

analysis for a postulated fuel handling 

accident inside containment during fuel 

loading and refueling operations. The 

proposed change does not involve any 

structure, system, or component relied upon 

to mitigate any design basis accident. The 

revised operations are consistent with the 

fuel handling accident analysis. 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 amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Previously approved analysis demonstrates 

that the resultant dose consequences are well 

within the appropriate acceptance criteria. 

The proposed change is bounded by the 

previously approved analysis, and thus the 

margin of safety, as defined by 10 CFR 50.67 

and Regulatory Guide 1.183, is maintained. 

Maintaining the capability to close the 

containment equipment hatch opening 

following an evacuation of the containment 

would further reduce the dose consequences 

in the event of a fuel handling accident 

inside containment and provides additional 

margin to the calculated doses. 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 

Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 70715

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: Thomas S. 

O’Neill, Associate General Counsel, 

AmerGen Energy Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Section Chief: Richard J. Laufer. 

Calvert Cliffs Nuclear Power Plant, Inc., 

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

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of amendment request: July 13, 

2004. 

Description of amendment request:

The proposed amendment would revise 

the fire protection license condition 

consistent with the guidance provided 

in Generic Letter 88–12, ‘‘Removal of 

Fire Protection Requirements from 

Technical Specifications.’’

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. Would not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

The proposed license amendment is an 

administrative change. The proposed change 

will revise the fire protection license 

condition consistent with the guidance 

provided in Generic Letter 88–12. This 

revision to the fire protection license 

condition was to be made at the time the fire 

protection requirements were relocated from 

the Technical Specifications to licensee 

controlled documents. However, this change 

was not requested, nor granted in License 

Amendment Request dated December 4, 

1996, approved in Amendment Nos. 227 and 

201. Therefore, the necessary change was not 

reflected in the Operating Licenses. 

This administrative request does not 

impact the probability or consequences of an 

accident previously evaluated. The 

incorporation of the requested change 

requires that an evaluation be performed to 

determine the need for prior NRC approval 

for changes to the Fire Protection Program. 

Changes to administrative programs will 

result from the addition of this condition in 

the Operating License. However, no changes 

to the facility or the way it is operated are 

expected to result from this change. 

Therefore, this proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Would not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

The proposed change is administrative. 

This request does not involve a change in the 

operation of the plant, and no new accident 

initiation mechanism is created by the 

proposed change, nor does the change 

involve a physical alteration of the plant. 

Therefore, the proposed change does not 

create the possibility of a new or different 

[kind] of accident from any accident 

previously evaluated. 

3. Would not involve a significant 

reduction in [a] margin of safety. 

The fire protection requirements were 

removed from the Technical Specifications in 

accordance with Generic Letter 88–12 in 

Amendment Nos. 227 and 201, with the 

exception of the change to the Operating 

License’s fire protection license condition. 

The proposed administrative change will 

require an evaluation be performed to 

determine the need for prior NRC approval 

for changes to the Fire Protection Program. 

No margin of safety is impacted by the 

proposed administrative change. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: James M. Petro, 

Jr., Esquire, Counsel, Constellation 

Energy Group, Inc., 750 East Pratt Street, 

5th floor, Baltimore, MD 21202. 

NRC Section Chief: Richard J. Laufer. 

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 7, 2004. 

Description of amendment request:

The proposed changes would allow 

performance of testing for nozzle 

containment spray blockage to be based 

on the occurrence of activities that 

could cause nozzle blockage rather than 

a fixed periodic basis. Currently, the 

testing for nozzle blockage is performed 

every 10 years and Dominion Nuclear 

Connecticut, Inc. (DNC) proposes to 

change this frequency to ‘‘following 

maintenance that could cause nozzle 

blockage’’. In addition, specific details 

limiting the testing method to an air or 

smoke test that are currently part of the 

surveillance requirements would be 

removed. The Technical Specification 

Bases section would be updated with 

applicable spray nozzle testing 

information and will be expanded to 

include visual inspection. 

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 spray nozzles and the associated 

containment spray systems are designed to 

perform accident mitigation functions only. 

The QSS [quench spray system] and RSS 

[recirculation spray system] and associated 

components are not considered as initiators 

of any analyzed accidents. The proposed 

change does not modify any plant equipment 

and only changes the frequency for 

performance of a surveillance test which 

does not impact any failure modes that could 

lead to an accident. Removing the testing 

details from the surveillance does not change 

the ability of the spray nozzles to function as 

assumed and therefore there is no affect on 

the consequence of any accident. Also the 

proposed change does not impact the 

capability of the QSS and RSS to perform 

accident mitigation functions and therefore 

does not impact the consequences of an 

accident. 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 QSS and RSS are not being physically 

modified and there is no impact on the 

capability of the systems to perform accident 

mitigation functions. No system setpoints are 

being modified and no changes are being 

made to the method in which borated water 

is delivered to the spray nozzles. The testing 

requirements imposed by this proposed 

change to check for nozzle blockage 

following activities that could cause nozzle 

blockage do not introduce new failure modes 

for the system. By removing the testing 

details from the surveillance requirement, 

additional flexibility in the testing 

methodology is allowed for verifying the 

nozzles are unobstructed and assists in 

ensuring operability of the systems. The 

proposed amendment does not introduce 

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 does not change or 

introduce any new setpoints at which 

mitigating functions are initiated. No changes 

to the design parameters of the QSS and RSS 

are being proposed. No changes in system 

operation are being proposed by this change 

that would impact an established safety 

margin. The proposed change modifies the 

frequency for verification of nozzle 

operability in such a way that continued high 

confidence exists for the containment spray 

systems to functions as designed. In addition, 

removing specific testing details from the 

surveillance does not affect the ability of the 

70716 Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 

spray nozzles to function as designed. 

Therefore, based on the above, the proposed 

amendment does not involve a significant 

reduction in a margin of safety. 

In summary, DNC concludes that the 

proposed amendment does not represent a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c).

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: Daniel S. Collins, 

Acting Section Chief. 

Entergy Nuclear Operations, Docket No. 

50–247, Indian Point Nuclear 

Generating Unit No. 2, Westchester 

County, New York 

Date of amendment request:

November 1, 2004. 

Description of amendment request:

The proposed amendment would allow 

the use of a new gantry crane as part of 

the cask handling system in the fuel 

storage building (FSB) for moving spent 

fuel casks up to 110 tons into and out 

of the spent fuel pit. 

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 amendment would allow 

the use of the new [IP2] FSB gantry crane for 

loads up to 110 tons, and the new crane will 

prevent the load from being dropped given a 

single malfunction or failure of a portion of 

the crane. The handling of a loaded spent 

fuel cask is below the maximum load that the 

crane is designed to handle. 

This change does not increase the 

probability of an accident previously 

evaluated because the probability of a load 

drop is eliminated. The new crane system is 

designed in accordance with NUREG–0554 

and Ederer’s Generic Licensing Topical 

Report EDR–1 (NP)–A, that if a portion of the 

crane lifting devices malfunctions or fails, 

the load will move a limited distance 

downward prior to backup restraints 

becoming engaged. The change does not 

increase the consequences of an accident. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would not involve a significant increase in 

the probability or consequences of an 

accident previously evaluated. 

2. The process for transporting a cask with 

the new crane is limited from the FSB truck 

bay floor to the cask pit area of the spent fuel 

pool. Once a cask is loaded with spent fuel, 

it is lifted from the spent fuel pit, and 

lowered into the truck bay. The cask is never 

carried over spent fuel in the spent fuel pit. 

Therefore, the change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. The [IP2] FSB gantry crane has been 

installed to comply with the single-failureproof requirements of NUREG–0554 and 

NRC-approved Ederer Topical Report EDR–1, 

Revision 3, dated October 8, 1982. The 

installation provides additional load carrying 

capability up to 110 tons and additional 

safety features to prevent a cask drop. The 

safety margins provided by the new crane 

prevent failure of the crane or any lifting 

devices associated with it. The 

implementation of NUREG–0612 general 

guidelines for the FSB gantry crane provides 

further assurance that safe load paths, 

procedures, crane operator training, and 

crane inspection and maintenance activities 

will be established to ensure crane operation 

is performed in a consistently reliable 

manner. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would not involve a significant reduction in 

[a] margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Section Chief: Richard J. Laufer. 

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: October 

29, 2004. 

Description of amendment request:

The proposed change to the Technical 

Specification (TS) assures that sufficient 

fuel oil inventories are available in the 

Emergency Diesel Generator (DG) Fuel 

Oil Storage Tank (FOST) to support the 

Extended Power Uprate (EPU) 

consistent with the current licensing 

basis. 

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, which accounts for 

the fuel oil consumption related to the EPU, 

will revise the minimum TS volumes 

associated with the DG FOST. The change 

continues to assure that each DG can provide 

on-site power in the event of an accident and 

thereby assist in the mitigation of the 

accident. 

The proposed change to the five day full 

load fuel oil volume results in a usable 

volume 37,000 gallons of fuel oil. The 

proposed change removes the unusable 

volume (760 gallons) and other conservatism 

(240 gallons) that were included in the 

current TS. The fuel oil volume continues to 

allow for a runtime of 5 days at full load with 

the removal of this conservatism. 

These changes will not affect the capability 

of the AC [alternate current] Sources to 

power the systems required to safely 

shutdown the plant. The proposed changes 

are not accident initiators nor do they 

adversely affect accident initiators or 

precursors. These changes do not affect the 

mitigation of any accident nor do they 

adversely affect structures, systems, or 

components that are utilized for the 

mitigation of any analyzed events. The 

proposed changes will have no affect on the 

radiological consequences of any accident. 

The proposed changes do not affect the 

source term, containment isolation, or 

radiological release assumptions used in the 

evaluation of radiological consequences. 

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. 

Fuel oil is not an accident initiator. 

Therefore, the possibility of a new or 

different kind of accident will not be created 

in relationship to the proposed changes to 

the TS. No modifications are proposed to the 

existing fuel oil storage system that would 

alter the design function or the ability of the 

DG to perform its safety function. 

Therefore, the proposed change will not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change to the 7-day time 

dependent fuel oil volume results in an 

increase in volume to accommodate fuel oil 

consumption needed to support the EPU. 

The reduced volume associated with the 5-

day full load volume is equivalent to less 

than one hour of runtime and does not result 

in a significant reduction in a margin of 

safety because the calculational method 

results in a conservative estimate of the 

amount of fuel that would be needed during 

a design bases accident. 

The proposed change does not result in a 

change of the design bases for the DG or its 

support systems. The system will continue to 

provide a reliable source of power for safe 

shutdown of the reactor, assuming the single 

failure of one of the DGs. Independence, 

redundancy, and testability are maintained 

such that the required safety function can be 

performed by either DG train. 

Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 70717

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: N.S. Reynolds, 

Esquire, Winston & Strawn, 1400 L 

Street NW., Washington, DC 20005–

3502. 

NRC Section Chief: Michael K. Webb, 

Acting. 

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3 (Waterford 3), St. Charles Parish, 

Louisiana 

Date of amendment request:

November 5, 2004. 

Description of amendment request:

Waterford 3 Technical Specification 

(TS) currently requires all the Dry 

Cooling Tower (DCT) fans with cooling 

coils under the missile grating to be 

operable during a tornado watch. If one 

(or more) of these DCT fans is 

inoperable during a tornado watch, it is 

required to be restored to operable 

status within one hour or place the 

plant in Hot Standby within 6 hours. 

The purpose of this TS change is to 

allow the plant to take credit for the 

DCT fans that are not under the missile 

grating to meet the fan requirements 

specified in TS Table 3.7–3. In addition, 

the proposed change will delete the 

requirement to monitor ambient 

temperature conditions when the DCT 

fan is inoperable on an inoperable train 

of the Ultimate Heat Sink (UHS). 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change will delete the 

requirement to have all the DCT fans with 

cooling coils under the missile grating 

operable during a tornado watch. It has been 

determined (using tornado missile strike 

probability methodology—TORMIS) that the 

probability of damage to the DCT 

components not under the missile grating 

(fans, motors, associated conduits, electrical 

boxes, and cooling coils) is acceptably low. 

With respect to the probability of occurrence 

or the consequences of an accident 

previously analyzed in the FSAR [Final 

Safety Analysis Report], the possibility of a 

tornado reaching Waterford 3 and causing 

damage to plant systems, structures and 

components, including the DCT fans, is a 

design basis event considered in the FSAR. 

The probability of a tornado-generated 

missile strike on the DCT components was 

analyzed using the NRC [Nuclear Regulatory 

Commission] Staff approved probability 

method TORMIS. TORMIS showed that the 

change from essentially relying on DCT fans 

with cooling coils under the missile grating 

to relying on all operable DCT fans during a 

tornado watch is acceptable and represents 

an acceptably low probability of occurrence 

of tornado generated missile strikes on the 

DCTs. On this basis, the proposed change is 

not considered to constitute a significant 

increase in the probability of occurrence or 

the consequences of an accident. 

The proposed change to TS Action 3.7.4.d 

eliminates an unnecessary requirement, to 

determine ambient conditions and verify 

compliance with TS Table 3.7–3, when an 

Ultimate Heat Sink (UHS) fan is inoperable 

due to its associated train of UHS being 

inoperable. The determination of ambient 

temperature conditions and validation of the 

required number of fans based on the 

temperature will continue to be required 

when an UHS fan is inoperable and the 

associated train of UHS is operable. The UHS 

fans will not dissipate the required heat load 

when the associated train of UHS is 

inoperable, assuming the coincident ambient 

wet bulb temperature (78 °F) at the 

historically highest ambient dry bulb 

temperature (102 °F). This change represents 

a burden reduction and has no impact on 

plant safety. This change also does not 

impact the initiators or mitigation of any 

design basis event. 

The proposed revision to TS Table 3.7–3 

ensures consistency with the revisions to the 

TS Actions. This change is administrative 

and has no impact on the initiators or the 

mitigation of accidents previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change will delete the 

requirement to have all the DCT fans with 

cooling coils under the missile grating 

operable during a tornado watch. It has been 

determined that the probability of damage to 

the DCT components not under the missile 

grating is acceptably low. A tornado at 

Waterford 3 is a design basis event 

considered in the FSAR. Therefore, the 

change will not contribute to the possibility 

of or be the initiator for any new or different 

kind of accident, or occur coincident with 

any of the design basis accidents in the 

FSAR. The low probability threshold 

established for tornado missile damage to 

system components is consistent with these 

assumptions. 

The proposed change to TS Action 3.7.4.d 

eliminates an unnecessary requirement, to 

determine ambient conditions and verify 

compliance with TS Table 3.7–3, when an 

Ultimate Heat Sink (UHS) fan is inoperable 

due to its associated train of UHS being 

inoperable. The determination of ambient 

temperature conditions will continue to be 

required when an UHS fan is inoperable with 

the associated train of UHS operable. There 

are no plant modifications or design changes 

proposed. 

The proposed revision to TS Table 3.7–3 

ensures consistency with the revisions to the 

TS Actions. This is an administrative change. 

The above changes also do not have any 

impact on plant systems nor do they have 

any impact on the way plant systems are 

operated. Therefore, the proposed changes do 

not create the possibility of a new or different 

kind of accident. 

3. Does the proposed change involve a 

significant reduction in a margin of safety?

Response: No. 

The proposed changes do not involve a 

significant reduction in a margin of safety. 

The existing licensing basis for Waterford 3 

with respect to the design basis event of a 

tornado reaching the plant, generating 

missiles, and directing them toward the DCT 

components is to provide positive missile 

barriers. The basis for the proposed change 

recognizes that there is a low probability, 

below an established acceptance limit, that a 

tornado missile will strike DCT components. 

The change from essentially relying on DCT 

fans with cooling coils under the missile 

grating to relying on all operable DCT fans 

during a tornado watch is acceptable and 

represents an acceptably low probability of 

occurrence of tornado generated missile 

strikes on the DCTs. Therefore, this change 

is not considered to constitute a significant 

decrease in the margin of safety. 

The proposed change to TS Action 3.7.4.d 

eliminates an unnecessary requirement, to 

determine ambient conditions and verify 

compliance with TS Table 3.7–3, when an 

Ultimate Heat Sink (UHS) fan is inoperable 

due to its associated train of UHS being 

inoperable. The determination of ambient 

temperature conditions will continue to be 

required when an UHS fan is inoperable with 

the associated train of UHS operable. When 

the UHS is not available, the fans cannot 

dissipate the required heat load, assuming 

the coincident ambient wet bulb temperature 

(78 °F) at the historically highest ambient dry 

bulb temperature (102 °F). Therefore, it is not 

necessary to monitor ambient temperature 

and ensure the fan requirements of TS Table 

3.7–3 are met when the UHS train is 

inoperable. This change represents an 

operational burden reduction and has no 

impact on plant safety. 

The proposed revision to TS Table 3.7–3 

ensures consistency with the revisions to the 

TS Actions. These changes are administrative 

and have no impact on the operation of the 

plant, mitigation of analyzed events, or plant 

safety, 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, Entergy concludes that 

the proposed amendment presents no 

significant hazards consideration under the 

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

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified.

70718 Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 

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 1400 L 

Street NW., Washington, DC 20005–

3502. 

NRC Section Chief: Michael K. Webb, 

Acting. 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois; Docket 

Nos. 50–254 and 50–265, Quad Cities 

Nuclear Power Station, Units 1 and 2, 

Rock Island County, Illinois 

Date of application for amendment 

request: February 27, 2004, as 

supplemented by letter dated October 

11, 2004. 

Description of amendment request:

The proposed amendments would 

revise the Dresden Nuclear Power 

Station and Quad Cities Nuclear Power 

Station technical specifications (TS) to 

add the Oscillation Power Range 

Monitor (OPRM) instrumentation to the 

TS. 

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 proposed change has no impact on 

any of the existing neutron monitoring 

functions. It activates the OPRM scram 

function and updates the TS to add the 

OPRM-related functions. 

Activation of the OPRM scram function 

will replace the current methods that require 

operators to insert an immediate manual 

reactor scram in the reactor operating region 

where thermal hydraulic instabilities could 

potentially occur. While this region will 

continue to be avoided during normal 

operation, certain transients, such as a 

reduction in reactor recirculation flow, could 

place the reactor in this region. Operation in 

this region, with the OPRM instrumentation 

scram function activated would no longer 

require an immediate manual scram and thus 

may potentially cause a marginal increase in 

the probability of occurrence of an instability 

event. This potential increase in probability 

is acceptable because the OPRM function 

will automatically detect the instability 

condition and initiate a reactor scram before 

the Minimum Critical Power Ratio (MCPR) 

Safety Limit is reached. Consequences of the 

potential instability event are reduced 

because of the more reliable automatic 

detection and suppression of an instability 

event, and the elimination of dependence on 

the manual operator actions. Operators will 

continue to monitor for indications of 

thermal hydraulic instability when the 

reactor is operating in regions of potential 

instability as a backup to the OPRM 

instrumentation. 

The potential for spurious reactor scrams 

has been evaluated. Operating experience 

with the OPRM has not resulted in the 

generation of any spurious reactor scram 

signals. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability of consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

The proposed changes replace procedural 

actions that were established to avoid 

operating conditions where reactor 

instabilities might occur with an NRC 

approved automatic detect and suppress 

function (i.e., OPRM). 

Potential failures in the OPRM trip 

function could result in either failure to take 

the required mitigating action or an 

unintended reactor scram. These are the 

same potential effects of failure of the 

operator to take the appropriate action under 

the current procedural actions. The effects of 

failure of the OPRM equipment are limited to 

reduced or failed mitigation, but such failure 

cannot cause an instability event or other 

type of accident. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

The OPRM trip function is being 

implemented to automate the detection and 

subsequent suppression of an instability 

event prior to exceeding the MCPR Safety 

Limit. The OPRM trip provides a trip output 

of the same type as currently used for the 

APRM [Average Power Range Monitor]. Its 

failure modes and types are identical to those 

for the present APRM output. Since the 

MCPR Safety Limit will not be exceeded as 

a result of an instability event following 

implementation of the OPRM trip function, it 

is concluded that the proposed change does 

not reduce the margin of safety. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration.

Attorney for licensee: Mr. Thomas S. 

O’Neill, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Section Chief: Gene Y. Suh. 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

Station, Unit Nos. 1 and 2 (BVPS–1 and 

2), Beaver County, Pennsylvania 

Date of amendment request: October 

4, 2004. 

Description of amendment request:

The proposed change requests approval 

to apply the Westinghouse best-estimate 

loss-of-coolant accident (BELOCA) 

analysis methodology to Beaver Valley 

Power Station, Unit Nos. 1 and 2, and 

requests an amendment of the related 

Technical Specifications. The BELOCA 

methodology has previously been 

approved on a generic basis by the NRC 

as presented in Topical Report WCAP–

12945-P-A, Volume 1 (Revision 2) and 

Volumes 2 through 5 (Revision 1), 

‘‘Code Qualification Document for BestEstimate LOCA Analysis,’’ March 1998. 

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. No physical changes are required as a 

result of implementing best-estimate large 

break loss of coolant accident (LOCA) 

methodology and associated Technical 

Specification changes. The plant conditions 

used in the analysis are bounded by the 

design conditions for all equipment in the 

plant. Therefore, there will be no increase in 

the probability of a LOCA. The consequences 

of a LOCA are not being increased, since it 

is shown that the emergency core cooling 

system is designed so that its calculated 

cooling performance conforms to the criteria 

contained in 10 CFR 50.46, Paragraph b. No 

other accident is potentially affected by this 

change. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any previously analyzed? 

No. There are no physical changes being 

made to the Beaver Valley Power Station 

units. No new modes of plant operation are 

being introduced. The parameters used in the 

analysis are within the design limits of the 

existing plant equipment. All plant systems 

will perform as designed during the response 

to a potential accident. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

analyzed. 

3. Does the proposed amendment involve 

a significant reduction in the margin of 

safety? 

Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 70719

No. It has been shown that the 

methodology used in the analysis would 

more realistically describe the expected 

behavior of plant systems during a postulated 

LOCA. Uncertainties have been accounted for 

as required by 10 CFR 50.46. A sufficient 

number of LOCAs with different break sizes, 

different locations and other variations in 

properties are analyzed to provide assurance 

that the most severe postulated LOCAs are 

addressed. It has been shown by analysis that 

there is a high probability that all criteria 

contained in 10 CFR 50.46, Paragraph b are 

met. 

Therefore the proposed change does not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mary O’Reilly, 

FirstEnergy Nuclear Operating 

Company, FirstEnergy Corporation, 76 

South Main Street, Akron, OH 44308. 

NRC Section Chief: Richard J. Laufer. 

Nine Mile Point Nuclear Station, LLC, 

Docket No. 50–220, Nine Mile Point 

Nuclear Station Unit No. 1, Oswego 

County, New York 

Date of amendment request: October 

15, 2004. 

Description of amendment request:

The licensee proposed to revise Section 

1.7, regarding the definition of 

‘‘Instrument Channel Calibration,’’ of 

the Technical Specifications by 

incorporating the additional guidance 

for instrument channels containing 

resistance temperature detector (RTD) 

and thermocouple sensors provided by 

the ‘‘Standard Technical Specifications, 

General Electric Plants, BWR [BoilingWater Reactor]/4 Specifications,’’

NUREG–1433, Revision 3. The revised 

definition would permit in place 

qualitative assessment of the RTDs and 

thermocouples, and to allow a signal to 

be injected downstream of the sensor for 

the purpose of calibrating the remainder 

of the channel. 

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 definition 

of Instrument Channel Calibration to allow 

RTD and thermocouple sensors to be 

qualitatively assessed with the remainder of 

the channel being calibrated normally. 

Instrument channel calibration is not an 

initiator of any accident previously 

evaluated. Furthermore, the proposed change 

will not affect the ability of the channel being 

calibrated to respond as assumed in any 

accident previously evaluated. The 

qualitative evaluation of sensor behavior for 

non-adjustable sensors will provide an 

accurate indication of sensor operation and 

will assure that portion of the channel is 

operating properly, ensuring that the 

consequences of an accident will remain as 

previously evaluated. Therefore, the 

proposed Technical Specification changes do 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated?

Response: No. 

The proposed change revises the definition 

of Instrument Channel Calibration to allow 

RTD and thermocouple sensors to be 

qualitatively assessed with the remainder of 

the channel being calibrated as at present. 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. The proposed change also 

does not adversely affect the operation or 

operability of existing plant equipment. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change involves the 

definition of Instrument Channel Calibration 

to allow RTD and thermocouple sensors to be 

qualitatively assessed with the remainder of 

the channel being calibrated normally. The 

proposed change to the Instrument Channel 

Calibration definition does not alter the 

ability of a channel to respond as designed 

or as assumed in the safety analyses. 

Therefore, this change does not involve a 

significant reduction in a margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mark J. 

Wetterhahn, Esquire, Winston & Strawn, 

1400 L Street, NW., Washington, DC 

20005–3502. 

NRC Section Chief: Richard J. Laufer. 

Nine Mile Point Nuclear Station, LLC, 

Docket No. 50–220, Nine Mile Point 

Nuclear Station Unit No. 1, Oswego 

County, New York 

Date of amendment request: October 

22, 2004. 

Description of amendment request:

The licensee proposed to revise the 

Technical Specifications (TSs), Section 

5.0, ‘‘Design Features,’’ by relocating the 

information to the Updated Final Safety 

Analysis Report (UFSAR). Specifically, 

the amendment would relocate these 

Sections: 5.3, ‘‘Reactor Vessel,’’ 5.4, 

‘‘Containment,’’ and 5.6, ‘‘Seismic 

Design.’’ The licensee stated that such 

information does not meet the criteria of 

10 CFR 50.36(c)(4) for inclusion in the 

TSs. The information to be relocated to 

the UFSAR already exists in the UFSAR, 

and will continue to be controlled by 10 

CFR 50.59 and 10 CFR 50.71(e). 

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 relocates certain 

design details from the TS to the UFSAR, 

where the information already exists. The 

UFSAR is maintained in accordance with 10 

CFR 50.71(e). Any future change to these 

design details as described in the UFSAR 

will be evaluated per the requirements of 10 

CFR 50.59 to assure that the change does not 

result in more than a minimal increase in the 

probability or consequences of an accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. The proposed change will 

not impose or eliminate any requirements, 

and adequate control of the information will 

be maintained in accordance with applicable 

regulatory requirements. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change has no impact on any 

analysis assumptions. The design details that 

70720 Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 

are being removed from the TS already exist 

in the UFSAR. Any future change to these 

design details described in the UFSAR will 

be evaluated per the requirements of 10 CFR 

50.59 to assure that the change does not 

result in a design basis limit [or] a fission 

product barrier being exceeded or altered. 

Therefore, the proposed change does not 

involve a significant reduction in [a] margin 

of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mark J. 

Wetterhahn, Esquire, Winston & Strawn, 

1400 L Street, NW., Washington, DC 

20005–3502. 

NRC Section Chief: Richard J. Laufer. 

Nuclear Management Company, LLC, 

Docket No. 50–305, Kewaunee Nuclear 

Power Plant, Kewaunee County, 

Wisconsin 

Date of amendment request: October 

14, 2004. 

Description of amendment request:

The proposed changes correct 

administrative errors in Technical 

Specifications 3.10.i and 6.9.a.4.A. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated?

NMC [Nuclear Management Company, the 

licensee] Response for Proposed Change to 

TS 3.10.i: No. The NMC has reviewed the 

proposed change in accordance with the 

provisions of 10 CFR 50.92 to show no 

significant hazards exist. This change is 

being proposed to correct an administrative 

error that currently exists within the KNPP 

[Kewaunee Nuclear Power Plant] Technical 

Specifications; therefore it would not have an 

affect on the probability of an accident 

previously evaluated. 

NMC Response for Proposed Change to TS 

6.9.a.4.A: No. The NMC has reviewed the 

proposed change in accordance with the 

provisions of 10 CFR 50.92 to show no 

significant hazards exist. This change is 

being proposed to correct an administrative 

error that currently exists within the KNPP 

Technical Specifications; therefore it would 

not have an affect on the probability of an 

accident previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

NMC Response for Proposed Change to TS 

3.10.i: No. The proposed change does not 

alter plant configuration, operating setpoints, 

or overall plant performance. Therefore, the 

proposed change would not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

NMC Response for Proposed Change to TS 

6.9.a.4.A: No. The proposed change does not 

alter plant configuration, operating setpoints, 

or overall plant performance. Therefore, the 

proposed change would not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

NMC Response for Proposed Change to TS 

3.10.i: No. The proposed change does not 

involve a significant reduction in a margin of 

safety. Inclusion of the omitted word ‘‘and’’

in TS 3.10.i will enhance the margin of 

safety. 

NMC Response for Proposed Change to 

6.9.a.4.A: No. The proposed change does not 

involve a significant reduction in a margin of 

safety. Correction of the references in TS 

Section 6.9.a.4.A will enhance 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: Bradley D. 

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

Box 1497, Madison, WI 53701–1497. 

NRC Section Chief: L. Raghavan. 

Pacific Gas and Electric Company, 

Docket No. 50–133, Humboldt Bay 

Power Plant, Unit 3, Humboldt County, 

California 

Date of amendment request: July 9, 

2004. 

Description of amendment request:

The Humboldt Bay Power Plant (HBPP), 

Unit 3, is a decommissioning nuclear 

power plant that was permanently 

shutdown in July 1976. In December of 

2003, Pacific Gas and Electric (PG&E or 

the licensee) applied for a license to 

store its spent fuel in an onsite dry cask 

independent spent fuel storage 

installation (ISFSI). Moving the spent 

fuel to an ISFSI would permit the 

licensee to begin significant 

decommissioning activities. The 

licensee has chosen to use a Holtec HI–

STAR HB spent fuel cask handling 

system involving a spent fuel 

multipurpose canister and overpack. To 

facilitate spent fuel transfer from the 

HBPP spent fuel pool to the ISFSI, the 

licensee will also need to install a new 

crane that can be used to lift the cask 

handling system loaded with spent fuel 

assemblies. The licensee states it will be 

able to satisfy the applicable guidance of 

NUREG–0612, ‘‘Control of Heavy Loads 

at Nuclear Power Plants,’’ and NUREG–

0554. ‘‘Single-Failure Proof Cranes for 

Nuclear Power Plants,’’ in performing 

the necessary movement of the HBPP 

spent fuel to dry cask storage. The 

licensee has requested a license 

amendment that approves the use of the 

crane and associated changes to the 

HBPP Defueled Safety Analysis Report 

(DSAR) along with analyses, design, and 

procedural changes required to 

implement transfer of the spent fuel 

from the spent fuel pool to the ISFSI. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. With the HI–STAR HB System and the 

associated design and handling procedures, 

all cask drops and other events, which could 

damage other spent fuel, have been 

precluded through the robust handling 

systems, and mechanical arrangement that 

preclude crane movement over spent fuel, 

meeting the guidelines of NUREG–0612. 

Revisions of the HBPP procedures 

implementing the control of heavy loads 

ensures that PG&E will meet the NUREG–

0612 guidelines and will protect the fuel 

storage locations and the new HI–STAR HB 

System loading/unloading activities. As a 

result of this design approach, a caskhandling accident that results in a significant 

offsite radiological release is not considered 

credible as demonstrated by the probabilistic 

evaluation that was performed using the 

guidelines of NUREG–0612 Appendix B and 

updated information from NUREG–1774 [‘‘A 

Survey of Crane Operating Experience at U.S. 

Nuclear Power Plants from 1968 through 

2002.’’] 

Other HBPP licensing-basis events, such as 

the drop of a spent fuel assembly, have not 

been affected by these changes and remain 

bounding events for potential radiological 

consequences. 

The proposed design of the dry cask 

system, the handling system, and associated 

procedural controls provide assurance that: 

(1) Operational errors and mishandling 

events, and (2) support system malfunctions 

will not result in an increase in the 

probability or consequence of an accident 

previously analyzed. 

The proposed changes to use the Holtec 

HI–STAR HB system have been evaluated for 

seismic events and tornado missile impacts 

and it has been determined that these 

changes will not result in an increase in the 

probability or consequences of an accident 

previously evaluated. The Fire Protection 

Program will ensure that the combustible 

materials are properly controlled such that 

the total combustibles meet the current 

program commitments. Therefore, the 

Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 70721

proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident.

2. Does the proposed amendment create 

the possibility of a new or different type of 

accident from any accident previously 

evaluated? 

No. The engineering design measures and 

the handling procedures preclude the 

possibility of new or different kinds of 

accidents. Damage to 10 CFR 50 structures, 

systems, and components from the cask 

handling and associated activities, and 

events resulting from possible damage to 

contained fuel have been considered. Both 

the types of accidents and the results remain 

within the envelope of existing HBPP DSAR 

licensing basis analyses, as demonstrated by 

the PG&E and Holtec analyses. 

The rupture of multipurpose canister 

(MPC) dewatering, forced helium 

dehydration or related closure system lines 

or the malfunction of equipment during cask 

handling operations resulting in radiological 

consequences are bounded by the HBPP 

DSAR fuel-handling accident analysis. 

Other design considerations, such as spent 

fuel pool (SFP) thermal, water chemistry and 

clarity, criticality, and structural, were 

evaluated and determined not to introduce 

the possibility of a new or different kind of 

accident from any previously evaluated. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

No. With the Holtec HI–STAR HB System, 

and the associated design and handling 

procedures, cask drops and other events have 

been precluded through robust load handling 

systems, providing defense-in-depth as 

described in NUREG–0612. Cask tipovers, 

while not considered credible, are shown to 

be below the 60g limit, preventing damage to 

the contained fuel assemblies (and associated 

structures), and meeting the analysis 

guidelines of NUREG–0612. As the existing 

licensing basis assumes a nonmechanistic 

drop damaging the SFP and all fuel, the 

result of this design approach with the 

minimization of drops and the associated 

structural challenges assure the margin of 

safety has been maintained. 

Other HBPP licensing-basis events, such as 

the drop of a spent fuel assembly, have not 

been affected by these changes and remain 

bounding events. Revision of HBPP 

procedures implementing the control of 

heavy loads to incorporate the additional 

restrictions on heavy loads movement will 

not affect the procedures or methodology 

used and will, therefore, not affect margins. 

Adverse effects from seismic events and/or 

cask drops or tipovers have been evaluated, 

assuring that the fuel, MPC, and overpack 

remain within their design bases. Since 

design basis criteria are fully satisfied, there 

is no impact on the margin of safety. 

The Fire Protection Program will continue 

to ensure that the combustible materials are 

properly controlled such that the total 

combustibles meet the current program 

commitments. Thus, there are no significant 

reductions in margin of safety associated 

with these changes. 

Other design considerations, such as SFP 

thermal, water chemistry, criticality, and 

structural, were evaluated and determined to 

not involve a reduction in a margin of safety.

Based on the above evaluations, the 

licensee concludes that the activities 

associated with the above changes 

present no significant hazards 

consideration under the standards set 

forth in 10 CFR 50.92 and accordingly, 

a finding by the NRC of no significant 

hazards consideration is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Richard F. 

Locke, Esquire, Pacific Gas and Electric 

Company, P.O. Box 7442, San 

Francisco, California 94120. 

NRC Section Chief: Claudia Craig. 

PPL Susquehanna, LLC, Docket Nos. 

50–387 and 50–388, Susquehanna 

Steam Electric Station, Units 1 and 2, 

Luzerne County, Pennsylvania 

Date of amendment request:

September 8, 2004. 

Description of amendment request:

The proposed amendment would revise 

Technical Specification 3.1.8, ‘‘Scram 

Discharge Volume (SDV) Vent and Drain 

Valves,’’ to allow a vent or drain line 

with one inoperable valve to be isolated 

instead of requiring the valve to be 

restored to Operable status within 7 

days. 

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on February 24, 2003 (68 FR 

8637), on possible amendments to revise 

the action for one or more SDV vent or 

drain lines with an inoperable valve, 

including a model safety evaluation and 

model no significant hazards 

consideration (NSHC) determination, 

using the consolidated line-item 

improvement process. The NRC staff 

subsequently issued a notice of 

availability of the models for referencing 

in license amendment applications in 

the Federal Register on April 15, 2003 

(68 FR 18294). The licensee affirmed the 

applicability of the model NSHC 

determination in its application dated 

September 8, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

A change is proposed to allow the affected 

SDV vent and drain line to be isolated when 

there are one or more SDV vent or drain lines 

with one valve inoperable instead of 

requiring the valve to be restored to operable 

status within 7 days. With one SDV vent or 

drain valve inoperable in one or more lines, 

the isolation function would be maintained 

since the redundant valve in the affected line 

would perform its safety function of isolating 

the SDV. Following the completion of the 

required action, the isolation function is 

fulfilled since the associated line is isolated. 

The ability to vent and drain the SDV is 

maintained and controlled through 

administrative controls. This requirement 

assures the reactor protection system is not 

adversely affected by the inoperable valves. 

With the safety functions of the valves being 

maintained, the probability or consequences 

of an accident previously evaluated are not 

significantly increased. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Accident 

Previously Evaluated 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. Thus, this change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in [a] Margin 

of Safety

The proposed change ensures that the 

safety functions of the SDV vent and drain 

valves are fulfilled. The isolation function is 

maintained by redundant valves and by the 

required action to isolate the affected line. 

The ability to vent and drain the SDV is 

maintained through administrative controls. 

In addition, the reactor protection system 

will prevent filling of the SDV to the point 

that it has insufficient volume to accept a full 

scram. Maintaining the safety functions 

related to isolation of the SDV and insertion 

of control rods ensures that the proposed 

change does not involve a significant 

reduction in the margin of safety.

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179. 

NRC Section Chief: Richard J. Laufer. 

PPL Susquehanna, LLC, Docket No. 50–

388, Susquehanna Steam Electric 

Station, Unit 2, Luzerne County, 

Pennsylvania 

Date of amendment request:

September 22, 2004. 

Description of amendment request:

The proposed amendment would extend 

70722 Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 

the validity of the reactor pressure 

vessel (RPV) pressure-temperature (P–T) 

limit curves from May 1, 2005, to May 

1, 2006. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The evaluation for the Unit 2 P–T limit 

curves for 32 EFPYs [effective full-power 

years] was performed using the approved 

methodologies of 10 CFR [Part] 50 Appendix 

G and Code Case–640. The curves generated 

from these methods were approved as 

Amendment 174 (Ref. 1) and are currently in 

the Unit 2 TS. These curves ensure the P–T 

limits will not be exceeded during any phase 

of reactor operation. Resolution of the current 

industry issues related to fluence calculation 

methodology required PPL to limit 

applicability of the curves to May 1, 2005 for 

Unit 2. The proposed change does not alter 

any of the technical information shown on 

the present P–T curves. The change extends 

the expiration date for one year while 

maintaining the total accumulated exposure 

well below the 32 EFPY maximum exposure 

lifetime limit. Therefore, there is no increase 

in the probability or consequences of any 

previously evaluated accident as a result of 

this change. 

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 involves changing 

the expiration date on the Unit 2 P–T limit 

curves. The change does not affect the 

present operating margin in the P–T limit 

curves for inservice leakage and hydrostatic 

pressure testing, non-nuclear heatup and 

cooldown, and criticality. Operation in 

accordance with the present P–T curves, 

developed in accordance with the provisions 

of ASME Code [American Society of 

Mechanical Engineers Boiler and Pressure 

Vessel Code], Section XI, Appendix G; 10 

CFR [Part] 50 Appendix G, and ASME Code 

Case–640 provides adequate protection 

against a non-ductile-type fracture of the 

RPV. This proposed change does not create 

the possibility of any new or different [kind] 

of accident. The change extends the 

expiration date of the present P–T curves and 

does not result in any new or unanalyzed 

operation of any system or piece of 

equipment important to safety. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The technical information contained in the 

present P–T curves approved by Amendment 

174 (Ref. 1) is not affected by this change. 

Extending the expiration date of the curves 

from May 1, 2005 to May 1, 2006 will not 

reduce the margin of safety to RPV brittle 

fracture. 

Since the Unit 2 P–T curves have a 

maximum lifetime exposure of 32 EFPYs and 

the anticipated exposure by May 1, 2006 will 

be well below the maximum value, the 

margin of safety is not reduced as the result 

of this change in expiration date. Resolution 

of the current industry issues related to 

fluence calculation methodology requires 

PPL to limit applicability of the Unit 2 P–T 

curves to May 1, 2006.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179 

NRC Section Chief: Richard J. Laufer. 

Tennessee Valley Authority (TVA), 

Docket No. 50–390, Watts Bar Nuclear 

Plant, Unit 1, Rhea County, Tennessee 

Date of amendment request:

September 23, 2004. 

Description of amendment request:

The proposed amendment would revise 

the Technical Specifications (TSs) to 

require automatic starting of the 

auxiliary feedwater (AFW) pumps upon 

trip of the Turbine Driven Main 

Feedwater (TDMFW) pumps only when 

one or more of TDMFW pumps are 

operating. 

Basis for proposed no significant 

hazards consideration determination:

As required by Title 10 of the Code of 

Federal Regulations (10 CFR), Section 

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 design basis events which impose 

AFW safety function requirements are loss of 

normal main feedwater, main feedline or 

main steamline break, loss of offsite power, 

loss of coolant accident, and small break loss 

of coolant accident. These accident 

evaluations assume actuation of AFW 

occurring due to low-low steam generator 

level or a safety injection signal. These 

signals are required safety related features 

unlike start-up of the AFW pumps due to the 

trip of both TDMFW pumps which is an 

anticipatory function and not required for 

either transient or accident analyses. 

Requiring this function only when the 

TDMFW pumps are running will not impact 

any previously evaluated design basis events. 

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. This TS change involves the automatic 

start of the AFW pumps when the TDMFW 

Pumps trip. This change involves a function 

that is not a safety related feature and, 

therefore, is not credited in either transient 

or accident analyses. Since this change only 

affects the point at which this trip function 

needs to be operable and does not affect the 

function that actuates AFW due to low-low 

steam generator level or a safety injection 

signal, it will not be an initiator to a new or 

different kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in margin of safety? 

No. This TS change involves the automatic 

start of the AFW pumps when the TDMFW 

pumps trip which is not a safety related plant 

function. This change does not change any 

values or limits involved in a safety related 

function. 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. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application request: October 

27, 2004. 

Description of amendment request:

The amendment would revise Technical 

Specification (TS) 3.7.3, ‘‘Main 

Feedwater Isolation Valves (MFIVs),’’ to 

add the main feedwater regulating 

valves (MFRVs) and the associated 

MFRV bypass valves (MFRVBVs). In 

addition, the allowed outage time, or 

completion time, for inoperable MFIVs 

would be extended. 

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 changes add the MFRVs and 

MFRVBVs to TS 3.7.3 and extend the 

Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 70723

Completion Time for one or more MFIVs 

inoperable from 4 hours to 72 hours. 

Extending the Completion Time is not an 

accident initiator and thus does not change 

the probability that an accident will occur. 

However, it could potentially affect the 

consequences of an accident if an accident 

occurred during the extended unavailability 

of the inoperable MFIV. The increase in time 

that the MFIV is unavailable is small and the 

probability of an event occurring during this 

time period which would require isolation of 

the MFW [main feedwater] flow paths is low. 

Moreover, the redundancy provided by the 

MFRVs and MFRVBVs, which have the same 

actuation signals and closure time 

requirements as the MFIVs, provides 

adequate assurance that automatic feedwater 

isolation will occur if called upon. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Closure of the MFIVs is required to 

mitigate the consequences of the Main Steam 

Line Break and Main Feedwater Line Break 

accidents. The MFRVs and MFRVBVs 

provide a diverse backup to this function. 

[The extended Completion Time for 

inoperable MFIVs is not an accident 

initiator.] The proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. The proposed change does not involve 

a significant reduction in a margin of safety. 

The proposed changes do not revise any 

Technical Specification [Safety] Limit or 

accident analysis assumption. Therefore, 

[they do] not involve a reduction in a margin 

of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: John O’Neill, 

Esq., Shaw, Pittman, Potts & 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037. 

NRC Section Chief: Robert A. Gramm. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application request: October 

27, 2004. 

Description of amendment request:

The amendment would delete or revise 

license conditions in the operating 

license for the Callaway Plant because 

the requirements are either obsolete or 

adequately described elsewhere. The 

amendment would also revise Technical 

Specification Tables 5.5.9–2, ‘‘Steam 

Generator Tube Inspection,’’ and 5.5.9–

3, ‘‘Steam Generator Repaired Tube 

Inspection,’’ to delete the requirement to 

notify the NRC pursuant to 10 CFR 

50.72(b)(2) if the steam generator tube 

inspection results in a C–3 classification 

because reporting requirements are 

given in the regulations. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

This request involves administrative 

changes only. The changes consist of 

duplicates or overly burdensome reporting 

requirements or the deletion of completed 

items required by [the TSs or] conditions 

from the original issuance of Operating 

License NPF–30 [for the Callaway Plant]. No 

actual plant equipment or accident analyses 

will be affected by the proposed changes. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated.

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

This request involves administrative 

changes only. The changes consist of 

duplicates or overly burdensome reporting 

requirements or the deletion of completed 

items required by [the TSs or] conditions 

from the original issuance of Operating 

License NPF–30. No actual plant equipment 

or accident analyses will be affected by the 

proposed change[s] and no failure modes not 

bounded by previously evaluated accidents 

will be created. Therefore, the proposed 

changes do not create 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. 

Margin of safety is associated with 

confidence in the ability of the fission 

product barriers (i.e., fuel and fuel cladding, 

Reactor Coolant System pressure boundary, 

and containment structure [pressure 

boundary]) to limit the level of radiation dose 

to the public. This request involves 

administrative changes only. 

No actual plant equipment or accident 

analyses will be affected by the proposed 

change[s]. The changes consist of duplicates 

or overly burdensome reporting requirements 

or the deletion of completed items required 

by [the TSs or] conditions from the original 

issuance of Operating License NPF–30. 

Additionally, the proposed changes will not 

relax any criteria used to establish safety 

limits, will not relax any safety system 

settings, or will not relax the bases for any 

limiting conditions of operation. Therefore, 

the proposed changes do not involve a 

significant reduction in the margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: John O’Neill, 

Esq., Shaw, Pittman, Potts & 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037. 

NRC Section Chief: Robert A. Gramm. 

Virginia Electric and Power Company, 

Docket Nos. 50–280 and 50–281, Surry 

Power Station, Unit Nos. 1 and 2, Surry 

County, Virginia 

Date of amendment request:

September 15, 2004. 

Description of amendment request:

The proposed changes will change the 

Administrative Controls Section of the 

Technical Specifications (TS) in order to 

incorporate title changes, change the 

location where the plant-specific titles 

and TS titles are correlated, and relocate 

the unit staff requirements to the 

Quality Assurance Program. These 

proposed changes will support the 

implementation of proposed Virginia 

Electric and Power Company Topical 

Report DOM–QA–1, ‘‘Nuclear Facility 

Quality Assurance Program 

Description,’’ currently under NRC staff 

review. In addition, these proposed TS 

changes eliminate the descriptions of 

the onsite and offsite safety review 

organizations. 

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. Operation of Surry Units 1 and 2 in 

accordance with the proposed license 

amendments would not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

The proposed change is administrative in 

nature and does not affect plant systems, 

structures or components (SSCs) or plant 

operation during normal or accident 

conditions. The proposed change only affects 

the designated titles of personnel, rewords or 

relocates requirements within TS or deletes 

requirements that are either not required to 

be part of TS or are already required by 

regulation. The change also relocates the 

detailed description of the onsite and offsite 

safety review organizations and non-licensed 

personnel qualification requirements to the 

Quality Assurance Program. Therefore, this 

change has no bearing on the probability of 

an accident. The management organizational 

structure and safety and operational reviews 

have not changed and, therefore, do not 

impact the ability of operating procedures or 

administrative controls to prevent or mitigate 

70724 Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 

a previously evaluated accident. As such, 

this change does not alter the conclusions of 

the existing safety analyses and therefore 

does not alter the consequences of an 

accident previously evaluated. 

2. Operation in accordance with the 

proposed license amendments would not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

The proposed administrative change 

continues to ensure that adequate 

management oversight exists at the plant in 

accordance with the existing Technical 

Specifications. The proposed change only 

affects the designated titles of personnel, 

rewords or relocates requirements within TS 

or deletes requirements that are either not 

required to be part of TS or are already 

required per regulation. The change also 

relocates the detailed description of the 

onsite and offsite safety review organizations 

and non-licensed personnel qualification 

requirements to the Quality Assurance 

Program. Therefore this change does not 

impact plant SSCs or plant operation and 

therefore does not create the possibility of an 

accident of a different type than evaluated 

previously. The management organizational 

structure and safety and operational reviews 

have not changed. Therefore, there is no 

change in the method of plant operation, 

operation review or system design review. 

There are no new or different accident 

scenarios, accident initiators, nor failure 

mechanisms that will be introduced due to 

this change. 

3. Operation in accordance with the 

proposed license amendments would not 

involve a significant reduction in a margin of 

safety. 

The proposed change only affects the 

designated titles of personnel, rewords or 

relocates requirements within TS or deletes 

requirements that are either not required to 

be part of TS or are already required per 

regulation. The change also relocates the 

detailed description of the onsite and offsite 

safety review organizations and non-licensed 

personnel qualification requirements to the 

Quality Assurance Program. Consequently, 

this change does not impact plant design, 

plant operation or any safety margin and, 

therefore, does not significantly reduce a 

margin of safety. 

This evaluation concludes that the 

proposed amendments to the Surry Units 1 

and 2 Technical Specifications do not 

involve a significant increase in the 

probability or consequences of a previously 

evaluated accident, do not create the 

possibility of a new or different kind of 

accident and 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. Lillian M. 

Cuoco, Esq., Senior Counsel, Dominion 

Resources Services, Inc., Millstone 

Power Station, Building 475, 5th Floor, 

Rope Ferry Road, Rt. 156, Waterford, 

Connecticut 06385. 

NRC Section Chief: Mary Jane RossLee (Acting).

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: July 23, 

2004. 

Description of amendment request:

The amendment would revise Technical 

Specification 3.6.3, ‘‘Containment 

Isolation Valves,’’ by (1) Adding the 

abbreviation ‘‘(CIV)’’ for containment 

isolation valve in Condition A of the 

Actions for the Limiting Condition for 

Operation; (2) deleting the Note and 

revising Condition A to be for only one 

penetration flow path with one CIV 

inoperable; (3) revising the completion 

time for Required Condition A.1 from 4 

hours to as much as 7 days depending 

on the category of the inoperable CIV; 

and (4) revising Condition C to be for 

two or more penetration flow paths with 

one CIV inoperable. The proposed 

amendment is based on Topical Report 

WCAP–15791–P, ‘‘Risk-Informed 

Evaluation of Extensions to 

Containment Isolation Valve 

Completion Times.’’

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 changes to the 

Completion Times do not change the 

response of the plant to any accidents and 

have an insignificant impact on the reliability 

of the containment isolation valves. The 

containment isolation valves will remain 

highly reliable and the proposed changes will 

not result in a significant increase in the risk 

of plant operation. This is demonstrated by 

showing that the impact on plant safety as 

measured by the large early release frequency 

(LERF) and incremental conditional large 

early release probabilities (ICLERP) is 

acceptable. These changes are consistent 

with the acceptance criteria in [the riskinformed] Regulatory Guides 1.174 and 

1.177. Therefore, since the containment 

isolation valves will continue to perform 

their [safety] functions with high reliability 

as originally assumed and the increase in risk 

as measured by LERF and ICLERP is 

acceptable, there will not be a significant 

increase in the consequences of any 

accidents. 

The proposed changes do not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, or 

configuration of the facility or the manner in 

which the plant is operated and maintained. 

The proposed changes do not alter or prevent 

the ability of structures, systems, and 

components (SSCs) from performing their 

intended [safety] function to mitigate the 

consequences of an initiating event within 

the assumed acceptance limits. The proposed 

changes do not affect the source term, 

containment isolation, or radiological release 

assumptions used in evaluating the 

radiological consequences of an accident 

previously evaluated. Further, the proposed 

changes do not increase the types or amounts 

of radioactive effluent that may be released 

offsite, nor significantly increase individual 

or cumulative occupational/public radiation 

exposures. The proposed changes are 

consistent with the safety analysis 

assumptions and resultant consequences [in 

Chapter 15, ‘‘Accident Analysis,’’ of the 

Updated Final Safety Analysis Report 

(USAR) for the plant]. 

Therefore, it is concluded that this change 

does not increase the probability of 

occurrence of a malfunction of equipment 

important to safety. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed changes do not result in a 

change in the manner in which the 

containment isolation valves provide plant 

protection. There are no design changes 

associated with the proposed changes. The 

changes to Completion Times do not change 

any existing accident scenarios, nor create 

any new or different accident scenarios. 

The changes do not involve a physical 

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

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. In addition, the changes do 

not impose any new or different 

requirements or eliminate any existing 

requirements. The changes do not alter 

assumptions made in the safety analysis. The 

proposed changes are consistent with the 

safety analysis assumptions and current plant 

operating practice. 

Therefore, the possibility of a new or 

different malfunction of safety related 

equipment is not created. 

3. The proposed change does not involve 

a significant reduction in a margin of safety. 

The proposed changes do not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not impacted by these 

changes. The proposed changes will not 

result in plant operation in a configuration 

outside the design basis. The calculated 

impact on risk is insignificant and is 

consistent with the acceptance criteria 

contained in Regulatory Guides 1.174 and 

1.177. 

Therefore, it is concluded that this change 

does not involve a significant reduction in 

the margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 70725

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: Robert Gramm. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity For a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

Tennessee Valley Authority, Docket No. 

50–390 Watts Bar Nuclear Plant, Unit 1, 

Rhea County, Tennessee 

Date of application for amendments:

October 29, 2004. 

Brief description of amendments:

Provide a one-time change to Function 

4a, ‘‘Reactor Coolant System (RCS) Hot 

Leg Temperature Indication,’’ of 

Technical Specification Table 3.3.4–1. 

This would allow continued operation 

until the next refueling outage (spring of 

2005) with one out of four RCS hot leg 

temperature indications inoperable in 

the Auxiliary Control Room. 

Date of publication of individual 

notice in the Federal Register:

November 5, 2004 (69 FR 64596). 

Expiration date of individual notice:

November 19, 2004. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated.

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room, 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 NRC 

Public Document Room (PDR) Reference 

staff at 1–800–397–4209, (301) 415–

4737 or by e-mail to pdr@nrc.gov. (Note:

Public access to ADAMS has been 

temporarily suspended so that security 

reviews of publicly available documents 

may be performed and potentially 

sensitive information removed. Please 

check the NRC Web site for updates on 

the resumption of ADAMS access.) 

AmerGen Energy Company, LLC, et al., 

Docket No. 50–219, Oyster Creek 

Nuclear Generating Station, Ocean 

County, New Jersey 

Date of application for amendment:

August 27, 2004, as supplemented by 

letters dated October 11 and 19, 2004. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications, Section 2.1.A, changing 

the safety limit minimum critical power 

ratio value from 1.09 to 1.10 for both 

four-or five-recirculation-loop 

operation, and from 1.10 to 1.12 for 

three-recirculation-loop operation. 

Date of Issuance: November 16, 2004. 

Effective date: November 16, 2004, 

and shall be implemented within 60 

days of issuance. 

Amendment No.: 252. 

Facility Operating License No. DPR–

16: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: September 14, 2004 (69 FR 

55467). The October 11 and 19, 2004, 

letters provided clarifying information 

within the scope of the original 

application and did not change the 

staff’s initial proposed no significant 

hazards consideration determination. 

The Commission’s related evaluation of 

this amendment is contained in a Safety 

Evaluation dated November 16, 2004. 

No significant hazards consideration 

comments received: No. 

AmerGen Energy Company, LLC, et. al., 

Docket No. 50–219, Oyster Creek 

Nuclear Generating Station (OCNGS), 

Ocean County, New Jersey, Docket No. 

50–289, Three Mile Island Nuclear 

Station, Unit 1 (TMI–1), Dauphin 

County, Pennsylvania 

Date of application for amendments:

March 23, 2004, as supplemented June 

16, 2004. 

Brief description of amendments: The 

amendments relocate the Independent 

Onsite Safety Review Group 

requirements from the Administrative 

Controls in Section 6 of the Technical 

Specifications to the Exelon Generation 

Company, LLC (EGC)/AmerGen Energy 

Company, LLC (AmerGen) Quality 

Assurance Topical Report (QATR) at 

TMI–1 and OCNGS. In addition, 

administrative corrections are included, 

which update references to the EGC/

AmerGen QATR, which has replaced 

the OCNGS and TMI–1 Operational 

Quality Assurance Plans. 

Date of issuance: November 8, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 251 and 252. 

Facility Operating License Nos. DPR–

16 and DPR–50: Amendments revised 

the Technical Specifications. 

Date of initial notices in Federal 

Register: May 11, 2004 (69 FR 26186). 

The supplement dated June 16, 2004, 

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 determinations. 

The Commission’s related evaluation 

of the amendments is contained in a 

70726 Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 

Safety Evaluation dated November 8, 

2004. 

No significant hazards consideration 

comments received: No. 

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:

April 23, 2004. 

Brief description of amendment: The 

amendment deletes Technical 

Specification Section 6.16, ‘‘PostAccident Sampling Programs NUREG 

0737 (II.B.3, II–F.1.2),’’ and the related 

requirements to maintain a Post 

Accident Sampling System. 

Date of issuance: November 22, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 180 days. 

Amendment No.: 253.

Facility Operating License No. DPR–

50. Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 31, 2004 (69 FR 

26187) 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated November 22, 

2004. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, 

LaSalle County Station, Units 1 and 2, 

LaSalle County, Illinois 

Date of application for amendments:

March 12, 2004, and supplemented by 

letters dated June 16 and September 2, 

2004. 

Brief description of amendments: The 

amendments modify the LaSalle 

Technical Specifications (TS) to 

eliminate selected response time testing 

requirements associated with Reactor 

Protection System instrumentation and 

Primary Containment Isolation 

instrumentation for Main Steam Line 

Isolation functions. Specifically, the 

changes revise the response time testing 

requirements for TS Section 3.3.1.1, 

‘‘Reactor Protection System (RPS) 

Instrumentation,’’ Reactor Vessel Steam 

Dome Pressure—High function and TS 

Section 3.3.6.1, ‘‘Primary Containment 

Isolation Instrumentation,’’ Reactor 

Vessel Water Level—Low Low Low, 

Level 1 and Main Steam Line Pressure—

Low functions. 

Date of issuance: November 19, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 169, 155. 

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments 

revised the TS. 

Date of initial notice in Federal 

Register: April 13, 2004 (69 FR 19569). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 19, 

2004. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket No. 50–255, Palisades Plant, 

Van Buren County, Michigan 

Date of application for amendment:

December 23, 2003. 

Brief description of amendment: The 

amendment modifies technical 

specification (TS) requirements to adopt 

the provisions of Industry/TS Task 

Force (TSTF) change TSTF–359, 

‘‘Increased Flexibility in Mode 

Restraints.’’

Date of issuance: November 10, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days. 

Amendment No.: 219. 

Facility Operating License No. DPR–

20. Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: September 16, 2004 (69 FR 

55844). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated November 10, 

2004. 

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: May 14, 

2004. 

Brief description of amendment: The 

amendment relocates the requirements 

of Technical Specification 3.3(1)a, 

‘‘Reactor Coolant System and Other 

Components Subject to ASME XI Boiler 

& Pressure Vessel Code Inspection and 

Testing Surveillance’’ and TS 3.4, 

‘‘Reactor Coolant System Integrity 

Testing,’’ to the Updated Safety 

Analysis Report (USAR). Requirements 

in TS 3.3(1)a were related to inservice 

inspection of ASME Class 1, 2, and 3 

components and requirements in TS 3.4 

were related to reactor coolant system 

integrity testing. 

Date of issuance: November 8, 2004. 

Effective date: November 8, 2004, and 

shall be implemented within 120 days 

from the date of issuance. 

Amendment No.: 230. 

Renewed Facility Operating License 

No. DPR–40: The amendment revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: June 22, 2004 (69 FR 34703) 

The Commission’s related evaluation 

of the amendment is contained in a 

safety evaluation dated November 8, 

2004. 

No significant hazards consideration 

comments received: No. 

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 application for amendments:

June 22, 2004, as supplemented on 

September 27, 2004. 

Brief description of amendments: The 

amendments revise the frequency 

associated with Surveillance 

Requirement (SR) 3.3.8.1.4, which 

directs the performance of the logic 

system functional test, from once every 

18 months to once every 24 months. The 

amendments change the SRs in Hatch, 

Units 1 and 2 Technical Specifications. 

Date of issuance: November 22, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance.

Amendment Nos.: 243/186. 

Renewed Facility Operating License 

Nos. DPR–57 and NPF–5: Amendments 

revised the Surveillance Requirements 

in the Technical Specifications. 

Date of initial notice in Federal 

Register: August 3, 2004 (69 FR 46592). 

The supplement dated September 27, 

2004, provided clarifying information 

that did not change the scope of the 

June 22, 2004, application nor the initial 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 22, 

2004. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority (TVA), 

Docket No. 50–390, Watts Bar Nuclear 

Plant, Unit 1, Rhea County, Tennessee 

Date of amendment request: October 

29, 2004, as supplemented November 5, 

2004. 

Description of amendment request:

The amendment provides a one-time 

change to Function 4a, ‘‘Reactor Coolant 

System (RCS) Hot Leg Temperature 

Indication,’’ of Technical Specification 

(TS) Table 3.3.4–1 to allow continued 

operations until the next refueling 

outage with one out of four RCS Hot Leg 

Temperature Indications inoperable in 

the Auxiliary Control Room. 

Federal Register / Vol. 69, No. 234 / Tuesday, December 7, 2004 / Notices 70727

1This figure is based on the SIA Report on Office 

Salaries In the Securities Industry 2003 

(Compliance Manager) and includes 35% for 

overhead charges.

Date of Issuance: November 19, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

immediately upon receipt. 

Amendment No.: 53. 

Facility Operating License No. (NPF–

90): Amendment revised the Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): Yes. On 

November 5, 2004, the Commission 

issued a notice (69 FR 64596) that 

included the staff’s proposed 

determination that the amendment 

request involves no significant hazards 

consideration (NSHC). The notice 

provided an opportunity to submit 

comments on the Commission’s 

proposed NSHC determination. No 

comments have been received. The 

notice also provided an opportunity to 

request a hearing by November 19, 2004, 

but indicated that if the Commission 

makes a final NSHC determination, any 

such hearing would take place after 

issuance of the amendment. The 

supplement of November 5, 2004, is 

within the scope of that notice, and did 

not change the proposed no significant 

hazards consideration. 

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 November 

19, 2004. 

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.

Dated at Rockville, Maryland, this 29th day 

of November, 2004. 

For the Nuclear Regulatory Commission. 

Ledyard B. Marsh, 

Director, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation.

[FR Doc. 04–26606 Filed 12–6–04; 8:45 am] 

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE 

COMMISSION 

Proposed Collection; Comment 

Request 

Upon written request, copies available 

from: Securities and Exchange 

Commission, Office of Filings and 

Information Services, Washington, DC 

20549. 

Extension: Rule 17a–4; SEC File No. 

270–198; OMB Control No. 3235–0279. 

Notice is hereby given that pursuant 

to the Paperwork Reduction Act of 1995 

(44 U.S.C. 3501 et seq.), the Securities 

and Exchange Commission 

(‘‘Commission’’) is soliciting comments 

on the collection of information 

summarized below. The Commission 

plans to submit this existing collection 

of information to the Office of 

Management and Budget for extension 

and approval. 

Rule 17a–4 requires exchange 

members, brokers and dealers to 

preserve for prescribed periods of time 

certain records required to be made by 

Rule 17a–3. In addition, Rule 17a–4 

requires the preservation of records 

required to be made by other 

Commission rules and other kinds of 

records which firms make or receive in 

the ordinary course of business. These 

include, but are not limited to, bank 

statements, cancelled checks, bills 

receivable and payable, originals of 

communications, and descriptions of 

various transactions. Rule 17a–4 also 

permits broker-dealers to employ, under 

certain conditions, electronic storage 

media to maintain records required to 

be maintained under Rules 17a–3 and 

17a–4. 

There are approximately 6,900 active, 

registered broker-dealers. The staff 

estimates that the average amount of 

time necessary to preserve the books 

and records as required by Rule 17a–4 

is 254 hours per broker-dealer per year. 

Thus the staff estimates that the total 

compliance burden for 6,900 

respondents is 1,752,600 hours. 

The staff believes that compliance 

personnel would be charged with 

ensuring compliance with Commission 

regulation, including Rule 17a–4. The 

staff estimates that the hourly salary of 

a compliance manager is $50 per hour.1

Based upon these numbers, the total 

cost of compliance for 6,900 

respondents is approximately $87.63 

million (1,752,600 yearly hours x $50). 

The total burden hour decrease of 

128,661 results from the decrease in the 

number of respondents from 7,217 to 

6,900.

Written comments are invited on: (a) 

Whether the proposed collection of 

information is necessary for the proper 

performance of the functions of the 

agency, including whether the 

information shall have practical utility; 

(b) the accuracy of the agency’s estimate 

of the burden of the proposed collection 

of information; (c) ways to enhance the 

quality, utility, and clarity of the 

information to be collected; and (d) 

ways to minimize the burden of the 

collection of information on 

respondents, including through the use 

of automated collection techniques or 

other forms of information technology. 

Consideration will be given to 

comments and suggestions submitted in 

writing within 60 days of this 

publication. 

Please direct your written comments 

to R. Corey Booth, Director/Chief 

Information Officer, Office of 

Information Technology, Securities and 

Exchange Commission, 450 5th Street, 

NW., Washington, DC 20549.

Dated: November 29, 2004. 

Jill M. Peterson, 

Assistant Secretary.

[FR Doc. E4–3498 Filed 12–6–04; 8:45 am] 

BILLING CODE 8010–01–P

SECURITIES AND EXCHANGE 

COMMISSION 

Submission for OMB Review; 

Comment Request 

Upon Written Request, Copies Available 

From: Securities and Exchange 

Commission, Office of Filings and 

Information Services, Washington, DC 

20549.

Extension: Rule 17f–1(g); SEC File No. 270–

30; OMB Control No. 3235–0290

Notice is hereby given that, pursuant 

to the Paperwork Reduction Act of 1995 

(44 U.S.C. 3501 et seq.), the Securities 

and Exchange Commission 

(‘‘Commission’’) has submitted to the 

Office of Management and Budget a 

request for extension of the previously 

approved collection of information 

discussed below. 

• Rule 17f–1(g) Requirements for 

reporting and inquiry with respect to 

missing, lost, counterfeit, or stolen 

securities. 

Rule 17f–1(g), under the Securities 

Exchange Act of 1934 (‘‘Act’’), requires 

that all reporting institutions (i.e., every 

national securities exchange, member 

thereof, registered securities association, 

broker, dealer, municipal securities 

dealer, registered transfer agent, 

registered clearing agency, participant 

therein, member of the Federal Reserve 

System, and bank insured by the FDIC) 

maintain and preserve a number of 

documents related to their participation 

in the Lost and Stolen Securities 

Program (‘‘Program’’) under Rule 17f–1. 

The following documents must be kept 

in an easily accessible place for three 

years, according to paragraph (g): (1) 

Copies or all reports of theft or loss 

(Form X–17F–1A) filed with the 

Commission’s designee: (2) all 

agreements between reporting