68180 Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from October 29, 

2004, through November 12, 2004. The 

last biweekly notice was published on 

November 9, 2004 (69 FR 64984). 

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 

Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 68181

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/

requestor to relief. A petitioner/

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

A request for a hearing or a petition 

for leave to intervene must be filed by: 

(1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, Attention: Rulemaking and 

Adjudications Staff; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland, 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) E-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HEARINGDOCKET@NRC.GOV; or (4) 

facsimile transmission addressed to the 

Office of the Secretary, U.S. Nuclear 

Regulatory Commission, Washington, 

DC, Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

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

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer of 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(i)–(viii).

For further details with respect to this 

action, see the application for 

amendment which is available for 

public inspection at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the 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 change would revise 

Technical Specification (TS) Table 4.1–

1 functional testing surveillance interval 

from monthly to semi-annually for the 

following reactor protection system 

instrument channels: Table 4.1–1, Item 

No. 4, ‘‘Power Range Channel,’’ Item No. 

7, ‘‘Reactor Coolant Temperature 

Channel,’’ Item No. 8, ‘‘High Reactor 

Coolant Pressure Channel,’’ Item No. 9, 

‘‘Low Reactor Coolant Pressure 

Channel,’’ Item No. 10,’’ Flux-Reactor 

Coolant Flow Comparator,’’ Item No. 11, 

‘‘Reactor Coolant Pressure-Temperature 

Comparator,’’ Item No. 12, ‘‘Pump Flux 

Comparator,’’ Item No. 13, ‘‘High 

Reactor Building Pressure Channel,’’

Item No. 45, ‘‘Loss of Feedwater Reactor 

Trip,’’ and Item No. 46, ‘‘Turbine Trip/

Reactor Trip.’’ The TS Section 4.1 Bases 

would be revised to reflect the proposed 

change from monthly to semi-annually 

and to specify that one channel is being 

tested every 46 days on a continual 

sequential rotation, which is consistent 

with the calculations of BAW–10167A, 

Supplement 1, and associated Nuclear 

Regulatory Commission Safety 

Evaluation Report that indicate that the 

reactor protection system retains a high 

level of reliability for this test interval. 

The proposed change would also revise 

TS Table 4.1–1 functional testing 

surveillance interval from monthly to 

quarterly for the following reactor 

protection system reactor trip devices: 

Table 4.1–1, Item No. 1, ‘‘Protection 

Channel Coincidence Logic,’’ and Item 

No. 2, ‘‘Control Rod Drive Trip Breaker 

and Regulating Rod Power SCRs.’’ The 

TS Section 4.1 Bases would be revised 

to reflect the proposed change from 

monthly to quarterly testing and to 

specify that one channel is being tested 

every 23 days on a continual sequential 

rotation, which is consistent with the 

calculations of BAW–10167A, 

Supplement 3, February 1998, and the 

NRC SER for BAW–10167A, 

Supplement 3, dated January 7, 1998, 

that indicate that the reactor trip system 

retains a high level of reliability for this 

test interval. 

Basis for proposed valuated 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 reactor protection system monitors 

parameters related to safe operation and trips 

the reactor to protect the reactor core against 

fuel cladding damage. It also assists in 

protecting against reactor coolant system 

damage caused by high system pressure by 

limiting energy input to the system through 

reactor trip action. Therefore, this change has 

no impact on the probability of an accident 

previously evaluated. The results of the 

reliability analyses conducted in accordance 

68182 Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 

with NRC [Nuclear Regulatory Commission] 

approved methodology and criteria show that 

the test interval extension of the reactor 

protection system instrument channels and 

reactor trip devices is not a significant 

contributor to trip system unavailability or 

the risk of core damage. The reactor 

protection system instrument channel and 

reactor trip device functional test 

surveillance program will continue to ensure 

that the reactor protection system is capable 

of performing its intended safety function 

during a design basis accident. 

Therefore, this change has no effect on the 

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 involves the reactor 

protection system instrument channel and 

reactor trip device surveillance test interval, 

which is not, in and of itself, considered to 

be an accident initiator. Postulated failure of 

the reactor protection system instrument 

channel or reactor trip device to function is 

an analyzed condition and does not 

constitute a new or different kind of accident. 

The proposed change does not create any 

new failure modes not bounded by 

previously analyzed accidents. 

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 results of the reliability analysis 

conducted in accordance with NRC approved 

methodology and criteria show that the test 

interval extension of the reactor protection 

system instrument channels and reactor trip 

devices is not a significant contributor to trip 

system unavailability or the risk of core 

damage. The Technical Specifications will 

continue to require the reactor protection 

system trip setpoints to remain within the 

assumptions of the accident analysis and that 

adequate reliability of the reactor protection 

system trip devices is maintained, thus 

preserving existing margins of safety. 

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: Thomas S. 

O’Neill, Associate General Counsel, 

AmerGen Energy Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Section Chief: Richard J. Laufer. 

Carolina Power & Light Company, 

Docket No. 50–261, H. B. Robinson 

Steam Electric Plant, Unit No. 2, 

Darlington County, South Carolina 

Date of amendment request: August 

20, 2004. 

Description of amendment request:

The proposed amendment would revise 

the Allowable Values for the following 

Reactor Protection System (RPS) 

instrumentation functions: Intermediate 

Range Neutron Flux, Reactor Coolant 

Flow—Low, Steam Generator Water 

Level—Low Coincident with Steam 

Flow/Feedwater Flow Mismatch, and 

Intermediate Range Neutron Flux (P–6) 

Interlock. Additionally, these changes 

revise the Allowable Value for the 

Engineered Safety Feature Actuation 

System Instrumentation function for 

High Steam Flow in Two Steam Lines 

Coincident with Steam Line Pressure—

Low. Also the proposed amendment 

would delete an unnecessary footnote 

associated with the applicability for the 

Automatic Trip Logic RPS 

instrumentation function. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

The proposal to revise the Allowable 

Values for the affected reactor protection and 

engineered safety feature actuation functions 

was developed in accordance with the 

current setpoint methodology for HBRSEP 

[H. B. Robinson Steam Electric Plant], Unit 

No. 2, thus ensuring that the probability and 

consequences of previously evaluated 

accidents are not significantly increased. The 

proposed deletion of the unnecessary 

footnote associated with the Automatic Trip 

Logic reactor protection instrumentation 

function does not change the requirements 

for operability of this function. Therefore, the 

proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated, because the factors that are used 

to determine the probability and 

consequences of accidents are not being 

affected. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

The proposed changes will continue to 

ensure that the operability of the previously 

described functions will be appropriately 

maintained. No physical changes to the 

HBRSEP, Unit No. 2, systems, structures, or 

components are being implemented. There 

are no new or different accident initiators or 

sequences being created by the proposed 

Technical Specifications changes. Therefore, 

these changes do not create the possibility of 

a new or different kind of accident from any 

accident previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in the margin of safety? 

The proposed changes, as previously 

described, ensure that the margin of safety for 

the applicable fission product barriers that 

are protected by these functions will 

continue to be maintained. This conclusion 

is based on the use of a valid setpoint 

methodology for determining the Allowable 

Values for the reactor protection and 

engineered safety feature actuation functions. 

Therefore, these changes do not involve a 

significant reduction in the margin of safety. 

Based on the preceding discussion, the 

requested changes do not involve a 

significant hazards consideration.

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: Steven R. Carr, 

Associate General Counsel—Legal 

Department, Progress Energy Service 

Company, LLC, Post Office Box 1551, 

Raleigh, North Carolina 27602. 

NRC Section Chief: Michael L. 

Marshall. 

Duke Energy Corporation, et al., Docket 

Nos. 50–413 and 50–414, Catawba 

Nuclear Station, Units 1 and 2, York 

County, South Carolina; Docket Nos. 

50–369 and 50–370, McGuire Nuclear 

Station, Units 1and 2, Mecklenburg 

County, North Carolina; Docket Nos. 

50–269, 50–270, and 50–287, Oconee 

Nuclear Station, Units 1, 2, and 3, 

Oconee County, South Carolina 

Date of amendment request:

September 28, 2004. 

Description of amendment request:

The proposed amendment would delete 

Technical Specification (TS) 5.6.1, 

‘‘Occupational Radiation Exposure 

Report,’’ and TS 5.6.4, ‘‘Monthly 

Operating Reports.’’

The NRC staff issued a notice of 

availability of a model no significant 

hazards consideration (NSHC) 

determination for referencing in license 

amendment applications in the Federal 

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

The licensee affirmed the applicability 

of the model NSHC determination in its 

application dated September 28, 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:

Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 68183

Criterion 1—The proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

The proposed change eliminates the TS 

reporting requirements to provide a monthly 

operating report of shutdown experience and 

operating statistics if the equivalent data is 

submitted using an industry electronic 

database. It also eliminates the TS reporting 

requirement for an annual occupational 

radiation exposure report, which provides 

information beyond that specified in NRC 

regulations. The proposed change involves 

no changes to plant systems or accident 

analyses. As such, the change is 

administrative in nature and does not affect 

initiators of analyzed events or assumed 

mitigation of accidents or transients. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated.

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, add any new 

equipment, or require any existing 

equipment to be operated in a manner 

different from the present design. Therefore, 

the proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Criterion 3—The proposed change does not 

involve a significant reduction in a margin of 

safety? 

This is an administrative change to 

reporting requirements of plant operating 

information and occupational radiation 

exposure data, and has no effect on plant 

equipment, operating practices or safety 

analyses assumptions. For these reasons, the 

proposed change does not involve a 

significant reduction in the margin of safety.

Based upon the reasoning presented 

above, the requested change does not 

involve significant hazards 

consideration. 

Attorney for licensee: Ms. Lisa F. 

Vaughn, Legal Department (PB05E), 

Duke Energy Corporation, 422 South 

Church Street, Charlotte, North Carolina 

28201–1006. 

NRC Section Chief: Mary Jane RossLee, Acting. 

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 

5, 2004. 

Description of amendment request:

The proposed amendments would 

revise Technical Specifications (TSs) 3/

4.3.1, ‘‘Reactor Trip System 

Instrumentation,’’ and 3/4.3.2, 

‘‘Engineered Safety Feature Actuation 

System Instrumentation,’’ to modify 

steam generator (SG) level allowable 

value setpoints. The proposed changes 

address recent generic issues involving 

new SG level uncertainty considerations 

and margins associated with 

Westinghouse-designed SGs. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The SG water level-low-low setpoint 

and allowable value have been revised to 

address Westinghouse Nuclear Safety 

Advisory Letter NSAL–03–9 and other 

considerations on steam generator water level 

uncertainties. The revised setpoint and 

allowable value calculations continues to 

follow the setpoint methodology previously 

approved for BVPS Unit No. 1 and No. 2 

while addressing newly identified level 

uncertainty considerations. The proposed 

changes to the SG water level-low-low 

Allowable Value for BVPS Unit No. 1 and 

No. 2 and to the SG water level-high-high 

Allowable Value for BVPS Unit No. 2 

continue [to] maintain the validity of the 

safety analysis limits used in the safety 

analyses that credit the actuations based on 

SG water level. 

The proposed changes do not alter the 

causes for any accident described in the 

Updated Final Safety Analysis Report 

(UFSAR) that credit the SG water level 

setpoint actuations. Therefore, they do not 

involve a significant increase in the 

probability of an accident previously 

evaluated. 

The proposed changes do not alter the 

accident analyses that credit the SG water 

level-low-low setpoint actuation or the 

associated accident acceptance criteria. 

Therefore, they do not involve a significant 

increase in the consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. The SG water level-low-low setpoint 

and allowable value have been revised to 

address Westinghouse Nuclear Safety 

Advisory letter NSAL–03–9 and other 

considerations on steam generator water level 

uncertainties. Implementation of the 

proposed setpoint changes have no 

significant effect on either the configuration 

of the plant, or the manner in which the 

plant is operated. The proposed changes to 

the SG water level-low-low allowable value 

for BVPS Unit No. 1 and No. 2 and to the SG 

water level-high-high allowable value for 

BVPS Unit No. 2 continue to maintain the 

validity of the safety analysis limits used in 

the safety analyses that credit the actuations 

based on SG water level. 

Therefore, since the plant configuration is 

not adversely changed and the proposed 

changes do not alter the accident analyses 

that credit actuation based on SG water level, 

the proposed change does not create the 

possibility of a new or different [kind of] 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The Reactor Trip System and 

Engineered Safety Feature Actuation System 

setpoint analysis methodology and 

acceptance criteria provide the margin of 

safety. The SG water level-low-low and SG 

water level-high-high actuation setpoint and 

allowable value have been calculated using 

the same methodology as previously 

approved for the BVPS Unit No. 1 and No. 

2 while addressing newly identified 

considerations needed to protect the limits 

used in the safety analyses. The applicable 

safety analyses have been performed and 

show acceptable results. 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. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: October 

25, 2004. 

Description of amendment request:

The proposed amendment would revise 

Technical Specification 2.1.1.2 for the 

dual recirculation loop and single 

recirculation loop Safety Limit 

Minimum Critical Power Ratio 

(SLMCPR) values to reflect results of a 

cycle specific calculation.

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The probability of an evaluated accident is 

derived from the probabilities of the 

individual precursors to that accident. 

Changing the SLMCPR does not increase the 

probability of an evaluated accident. The 

change does not require any physical plant 

modifications, physically affect any plant 

components, or entail changes in plant 

68184 Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 

operation. Therefore, no individual 

precursors of an accident are affected. 

The consequences of an evaluated accident 

are determined by the operability of plant 

systems designed to mitigate those 

consequences. Limits have been established, 

consistent with NRC approved methods, to 

ensure that fuel performance during normal, 

transient, and accident conditions is 

acceptable. The proposed change 

conservatively establishes the safety limit for 

the minimum critical power ratio (SLMCPR) 

for Cooper Nuclear Station Cycle 23 such that 

the fuel is protected during normal operation 

and during any plant transients or 

anticipated operational occurrences. 

The proposed change revises the SLMCPR 

to protect the fuel during normal operation 

as well as during any transients or 

anticipated operational occurrences. 

Operational limits Minimum Critical Power 

Ratio (MCPR) are established based on the 

proposed SLMCPR to ensure that the 

SLMCPR is not violated during all modes of 

operation. This will ensure that the fuel 

design safety criteria (i.e., that at least 99.9% 

of the fuel rods do not experience transition 

boiling during normal operation and 

anticipated operational occurrences) is met. 

Since the operability of plant systems 

designed to mitigate any consequences of 

accidents has not changed, the consequences 

of an accident previously evaluated are not 

expected to increase. 

Based on the above NPPD [Nebraska Public 

Power District] concludes that the proposed 

changes do not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Creation of the possibility of a new or 

different kind of accident would require the 

creation of one or more new precursors of 

that accident. New accident precursors may 

be created by modifications of the plant 

configuration or changes in allowable modes 

of operation. The proposed change does not 

involve any modifications of the plant 

configuration or allowable modes of 

operation. The proposed change to the 

SLMCPR assures that safety criteria are 

maintained for Cycle 23. 

Based on the above NPPD concludes that 

the proposed changes do not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The value of the proposed SLMCPR 

provides a margin of safety by ensuring that 

no more than 0.1% of the rods are expected 

to be in boiling transition if the MCPR limit 

is not violated. The proposed change will 

ensure the appropriate level of fuel 

protection is maintained. Additionally, 

operational limits are established based on 

the proposed SLMCPR to ensure that the 

SLMCPR is not violated during all modes of 

operation. This will ensure that the fuel 

design safety criteria (i.e., that at least 99.9% 

of the fuel rods do not experience transition 

boiling during normal operation as well as 

anticipated operational occurrences) are met. 

Based on the above NPPD concludes that 

the proposed changes do not involve a 

significant reduction in a margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. John R. 

McPhail, Nebraska Public Power 

District, Post Office Box 499, Columbus, 

NE 68602–0499. 

NRC Acting Section Chief: Michael K. 

Webb. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey

Date of amendment request: April 15, 

2004. 

Description of amendment request:

The proposed change would modify the 

Salem Updated Final Safety Analysis 

Report (UFSAR) with respect to fire 

protection requirements for the 4160 

Volt Switchgear Rooms, 460 Volt 

Switchgear Rooms, and the Lower 

Electrical Penetration Area Rooms. 

Specifically, the amendment would 

reduce the UFSAR description of the 

Carbon Dioxide Tank volume from 

being able to provide two full discharges 

to an affected room to one full and one 

partial discharge to an affected room. 

Additionally, the assumed ability of the 

Carbon Dioxide system would be 

reduced from an ability to produce a 

CO2 concentration of 50% for 30 

minutes to an ability to produce a CO2

concentration of 27.6% for a length of 

time sufficient to suppress a fire and 

allow the PSEG Nuclear Fire 

Department to respond. 

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 likelihood of a fire event is not 

increased since the proposed change does not 

alter the fire hazards contained in the plant. 

The ability to achieve and maintain safe 

shutdown in the event of a fire is not 

impacted by the reduction of CO2

concentration, since the Fire Brigade will 

respond in ample time and extinguish a fire 

using alternate means. In addition, the 

proposed changes to the UFSAR would not 

change any response to a fire event. Also, the 

probability of occurrence or the 

consequences for an accident or malfunction 

of equipment is not increased by the 

proposed changes since the response to a fire 

event would not change and the fire brigade 

would continue to respond rapidly to any 

fires or fire alarms. Further, the proposed 

changes do not alter the way any structure, 

system, or component (SSC) functions, do 

not modify the manner in which the plant is 

operated, and do not significantly alter 

equipment out-of-service time. Changing the 

CO2 concentration requirement in the 4160 

Volt Switchgear Rooms, 460 Volt Switchgear 

Rooms and Lower Penetration Area Rooms at 

Salem Units 1 and 2 does not change the 

probability or consequences of any accident 

and dose consequences are unaffected. No 

changes to the design of structures, systems, 

or components (SSC) are made and there are 

no effects on accident mitigation. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequence 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 possibility of a new or different kind 

of accident from any accident or malfunction 

in the Salem Updated Final Safety Analysis 

Report (UFSAR) is not created. The design 

basis event applicable to the proposed 

change is a fire in the 4160 Volt Switchgear 

Rooms, 460 Volt Switchgear Rooms and 

Lower Penetration Area Rooms at Salem 

Units 1 and 2. Therefore a different accident 

is not created. In addition, the proposed 

changes cannot initiate an accident. Further, 

the proposed changes to the UFSAR do not 

change the design function or operation of 

any SSCs. 

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 change involve a significant 

reduction in a margin of safety? 

Response: No. 

The reduction in CO2 concentration 

provides ample response time for the onsite 

dedicated fire brigade to respond to a fire 

event and a 20% safety factor in CO2

concentration remains. The proposed 

changes do not affect the ability to safely 

shutdown and maintain the shutdown 

conditions of either unit following a fire in 

the affected areas. The proposed changes do 

not rely on compensatory measures or 

actions deviating from the licensing or design 

basis. In addition, the proposed changes do 

not change the margin of safety since no 

SSCs are changed. The results of accident 

analysis remain unchanged by the proposed 

changes to the UFSAR. 

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 

Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 68185

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration.

Attorney for licensee: Jeffrie J. Keenan, 

Esquire, Nuclear Business Unit—N21, 

P.O. Box 236, Hancocks Bridge, NJ 

08038. 

NRC Section Chief: James W. Clifford. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application request:

September 17, 2004. 

Description of amendment request:

The amendment is to support the 

replacement of the steam generators 

(SGs) at Callaway during the refueling 

outage in the Fall of 2005. The 

amendment would (1) change the 

affected technical specifications (TSs) 

such as the reactor core safety limits (TS 

2.1.1), reactor trip system (RTS) and 

engineered safety feature actuation 

system (ESFAS) instrumentation (TSs 

3.3.1 and 3.3.2), reactor coolant system 

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

3.4.5, 3.4.6, and 3.4.7), RCS operational 

leakage (TS 3.4.13), SG tube integrity 

(new TS 3.4.17), main steam safety 

valves (TS 3.7.1), SG surveillance 

program (TS 5.5.9), containment 

integrated leakage rate testing (ILRT) 

program (TS 5.5.16), and SG inspection 

report (TS 5.6.10); (2) revise the affected 

transient analyses such as excessive 

increase in secondary steam flow event, 

loss of normal feedwater event, transient 

mass and energy releases, radiological 

consequences of associated events, and 

containment pressure/temperature 

responses; and (3) revise nuclear steam 

and supply system (NSSS) design 

parameters and transients, and fatigue 

usage factors and stresses for the 

replacement SGs. The amendment 

involves the following areas of change 

to the license: nuclear steam supply 

system evaluations for the replacement 

steam generators, trip time delay (TTD) 

elimination for certain RTS and ESFAS 

functions, the SG surveillance program 

in Technical Specification Task Force 

(TSTF) No. 449 (TSTF–449), and the 

post-modification containment ILRT 

exception. 

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 for the above areas of 

review, which is presented below (with 

the terms defined in the plant Technical 

Specifications capitalized):

1. The proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Nuclear Steam Supply System Evaluations 

for Replacement Steam Generators 

As discussed in the NSSS Licensing Report 

(Appendix A to this amendment application), 

all acceptance criteria continue to be met. All 

major NSSS components (e.g., Reactor 

Vessel, Pressurizer, RCPs [(reactor coolant 

pumps)], Steam Generators, etc.) have been 

assessed with respect to bounding conditions 

expected for replacement steam generator 

(RSG) conditions. In all cases operation has 

been found to be acceptable. Major systems 

and subsystems (e.g., safety injection, RHR 

[residual heat removal], etc.) have been 

reviewed and acceptable performance has 

been verified for their normal operation and, 

as applicable, for their safety-related 

functions. All reactor trip and ESFAS 

actuation setpoints have been assessed, and 

the proposed setpoint modifications will 

assure adequate protection is afforded for all 

design basis events. 

The reactor core safety limits have been 

revised based on the RSG project parameters. 

All of the acceptance criteria for the accident 

analyses (e.g., DNBR [departure from 

nucleate boiling ratio] limits, fuel centerline 

temperatures, etc.) continue to be met with 

the revised safety limit lines. Therefore, the 

revised core safety limit line changes are 

acceptable. The proposed changes to the 

reactor core safety limits will not initiate any 

accidents; therefore, they do not increase the 

probability of an accident previously 

evaluated in the FSAR [Callaway Final Safety 

Analysis Report]. The comprehensive 

analytical efforts performed to support the 

proposed RSG conditions include a 

reanalysis or evaluation of all accident 

analyses that are impacted by the revised 

reactor core safety limits. 

The changes in various SG-related RTS and 

ESFAS Allowable Values have resulted from 

the analyses performed to support plant 

operation at the proposed RSG conditions. 

Setpoint uncertainty calculations confirm the 

acceptability of these revised Allowable 

Values. The affected RTS and ESFAS 

Allowable Values have been modified to 

reflect the results of updated setpoint 

calculations based on plant-specific 

uncertainties, calibration practices, 

calibration equipment, and installed 

hardware and procedures. The Allowable 

Values were calculated using the same 

Westinghouse setpoint methodology used for 

the current trip setpoints, but improved in a 

conservative fashion to include refinements 

that better reflect plant calibration practices 

and equipment performance. These 

refinements include the incorporation of a 

sensor reference accuracy term to address 

repeatability effects when performing a single 

pass calibration (i.e., one up and one down 

pass at several points verifies linearity and 

hysteresis, but not repeatability). In addition, 

sensor and rack error terms for calibration 

accuracy and drift are grouped in the 

Channel Statistical Allowance equation with 

their dependent measurement and test 

equipment (M&TE) terms, then combined 

with the other independent error terms using 

the square root sum of the squares (SRSS) 

methodology. This improved setpoint 

methodology has been previously review[ed] 

and approved by the NRC. The proposed RTS 

and ESFAS Allowable Value changes will not 

initiate any accidents; therefore, they do not 

increase the probability of an accident 

previously evaluated in the FSAR. The 

comprehensive analytical effort performed to 

support the proposed RSG conditions 

included a reanalysis or evaluation of all 

accident analyses that are impacted by the 

revised RTS and ESFAS Allowable Values. 

All systems will function as designed. 

The decrease in the Maximum Allowable 

Power for 3 OPERABLE MSSVs [main steam 

safety valves] per SG from < 49% of Rated 

Thermal Power to < 45% of Rated Thermal 

Power resulted from the analyses and 

evaluations performed to support plant 

operation at the proposed RSG conditions. 

The accident analysis acceptance criteria 

continue to be met with these changes. These 

proposed plant system changes do not 

increase the probability of an accident 

previously evaluated in the FSAR. The 

comprehensive analytical effort performed to 

support the proposed RSG conditions has 

included a review and evaluation of all 

components and systems (including interface 

systems and control systems) that could be 

affected by this change. All systems will 

function as designed. The change in the 

manner in which the Reactor Coolant Flow—

Low Allowable Value is defined (while 

retaining the same numerical value), the 

change in the manner in which RCS average 

temperature is defined and the reduced 

upper limit for nominal T-avg [average 

temperature] at full power conditions in the 

Overtemperature DT [delta temperature] and 

Overpower DT setpoint equations, and the 

changes to the pressurizer pressure and RCS 

average temperature limits in the DNB LCO 

[departure from nucleate boiling limiting 

condition for operation] [TS] 3.4.1 have also 

been evaluated. None of these proposed 

changes will initiate any accidents; therefore, 

the probability of an accident has not been 

increased. 

The potential dose consequences have 

been analyzed with respect to the above 

changes collectively. The dose increases are 

less than minimal (i.e., <10% of the margin 

between the regulatory limits and the 

currently reported doses). The applicable 

dose acceptance criteria continue to be met. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated.

Trip Time Delay Elimination 

This design change will eliminate only the 

Trip Time Delay portion of the SG Water 

Level Low-Low trip functions and return that 

portion of the design to condition that 

existed prior to Callaway Amendment 43 

dated April 14, 1989. The coincidence logic 

in the Solid State Protection System will be 

unaffected. In all other regards, the design of 

the RTS and ESFAS instrumentation will be 

unaffected. These protection systems will 

continue to function in a manner consistent 

with the plant design basis. All design, 

material, and construction standards that 

were applicable prior to this amendment 

request are maintained. 

68186 Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 

The probability and consequences of 

accidents previously evaluated in the FSAR 

are not adversely affected because the 

removal of the trip time delay circuitry 

assures a faster response by the affected trip 

functions, consistent with the safety analysis 

acceptance criteria and the original plant 

licensing basis. 

The proposed change will not affect the 

probability of any event initiators. There will 

be no degradation in the performance of, or 

an increase in the number of challenges 

imposed on, safety-related equipment 

assumed to function during an accident 

situation. There will be no change to normal 

plant operating parameters or accident 

mitigation performance. 

The proposed change will not alter any 

assumptions or change any mitigation actions 

in the radiological consequence evaluations 

in the FSAR. 

Therefore, the proposed TTD elimination 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

TSTF–449 Generic Licensing Change 

Package 

This proposed change requires a Steam 

Generator Program that includes performance 

criteria that will provide reasonable 

assurance that the steam generator (SG) 

tubing will retain integrity over the full range 

of operating conditions (including startup, 

operation in the power range, hot standby, 

cooldown, and all anticipated transients 

included in the design specification). The SG 

performance criteria are based on tube 

structural integrity, accident induced 

leakage, and operational LEAKAGE. 

A steam generator tube rupture (SGTR) 

event is one of the design basis accidents that 

are analyzed as part of a plant’s licensing 

basis. In the analysis cases for the SGTR 

event at Callaway Plant, a primary to 

secondary LEAKAGE rate of 1 gallon per 

minute (gpm) to the unaffected SGs is 

assumed, in excess of the RCS Operational 

LEAKAGE rate limit in TS 3.4.13, and the 

LEAKAGE rate associated with a doubleended rupture of a single tube in the 

ruptured SG is also assumed. For other 

design basis accidents such as main steam 

line break (MSLB), rod ejection, and reactor 

coolant pump locked rotor, the SG tubes are 

assumed to retain their structural integrity 

(i.e., they are assumed not to rupture). These 

additional analyses for Callaway Plant 

assume, as an initial condition, that primary 

to secondary LEAKAGE for all SGs is 1 gpm. 

The accident induced leakage criterion 

introduced by the proposed change to TS 

5.5.9 accounts for tubes that may leak during 

design basis accidents. The accident induced 

leakage criterion limits this leakage to no 

more than the 1 gpm value assumed in the 

accident analyses. 

The SG performance criteria added to TS 

5.5.9 identify the standards against which 

tube integrity is to be measured. Meeting the 

performance criteria provides reasonable 

assurance that the SG tubing will remain 

capable of fulfilling its specific safety 

function of maintaining reactor coolant 

pressure boundary integrity throughout each 

operating cycle and in the unlikely event of 

a design basis accident. The performance 

criteria are only a part of the Steam Generator 

Program required by the proposed change to 

TS 5.5.9. The program, defined by NEI 

[Nuclear Energy Institute] 97–06, Steam 

Generator Program Guidelines, includes a 

framework that incorporates a balance of 

prevention, inspection, evaluation, repair, 

and leakage monitoring. 

The consequences of design basis accidents 

are, in part, functions of the DOSE 

EQUIVALENT I–131 in the primary coolant 

and the primary to secondary LEAKAGE 

rates resulting from an accident. Therefore, 

limits are included in TS 3.4.13 for RCS 

Operational leakage and in TS 3.4.16 for 

DOSE EQUIVALENT I–131 in the primary 

coolant to ensure the plant is operated within 

its analyzed condition. The radiological 

consequence analyses at Callaway Plant 

assume that the primary to secondary 

LEAKAGE rate is 1 gpm (more conservative 

than the limit in TS 3.4.13), and that the 

reactor coolant activity levels of DOSE 

EQUIVALENT I–131 are at the TS 3.4.16 

limits. 

The proposed TSTF–449 changes reflect 

the design of the replacement SGs, but do not 

affect their method of operation or primary 

or secondary coolant chemistry controls. The 

proposed changes update the TS and 

enhance the requirements for SG inspections. 

The proposed changes do not adversely 

impact the conclusions of any previously 

evaluated design basis accident and are an 

improvement over the existing TS. 

Therefore, this proposed change to 

implement TSTF–449 does not affect the 

consequences of a SGTR accident and the 

probability of such an accident is reduced. In 

addition, this proposed change does not 

affect the consequences of an MSLB, rod 

ejection, reactor coolant pump locked rotor, 

or any other accident event involving the 

potential release of radioactive fluids from 

the secondary side of Callaway Plant. 

[Therefore, this proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated.] 

Post-Modification ILRT Exception 

This proposed change would provide 

Callaway Plant with an exception from 

performing a post-modification containment 

integrated leak rate test following the 

replacement of the steam generators during 

Refuel [Outage] 14. 

Integrated leak rate tests are performed to 

assure the leak-tightness of the primary 

containment boundary system, and as such 

they are not accident initiators. Therefore, 

not performing an integrated leak rate test 

will not affect the probability of an accident 

previously evaluated. The intent of postmodification integrated leak rate testing 

requirements is to assure the leak-tight 

integrity of the area affected by the 

modification. For the Callaway Plant steam 

generator replacement modification, this 

intent will be satisfied by performing the 

American Society of Mechanical Engineers 

code required inspections and tests. Since 

the leak-tightness integrity of the primary 

containment boundary affected by the steam 

generator replacement will be assured, there 

is no change in the containment boundary’s 

ability to confine radioactive materials 

during an accident. Therefore, adding a 

Technical Specification exception from the 

steam generator replacement postmodification integrated leak rate testing 

requirements does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

2. The proposed changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Nuclear Steam Supply System Evaluations 

for Replacement Steam Generators 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures are introduced as a result of 

this amendment. There will be no adverse 

effect or challenges imposed on any safetyrelated system as a result of this amendment. 

This amendment does not alter the safe 

performance of the plant protection systems 

to trip the reactor when necessary or actuate 

ESF [engineered safety feature] systems. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

Trip Time Delay Elimination 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures are introduced as a result of 

this amendment. There will be no adverse 

effect or challenges imposed on any safetyrelated system as a result of this amendment. 

This amendment does not alter the safe 

performance of the plant protection systems 

to trip the reactor when necessary or actuate 

ESF systems. 

Therefore, this proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

TSTF–449 Generic Licensing Change 

Package

The proposed performance based 

requirements are an improvement over the 

requirements imposed by the existing TS. 

Implementation of the proposed Steam 

Generator Program will not introduce any 

adverse changes to the plant design basis or 

postulated accidents resulting from potential 

tube degradation. The result of the 

implementation of the Steam Generator 

Program will be an enhancement of SG tube 

performance. Primary to secondary 

LEAKAGE that may be experienced during 

all plant conditions will be monitored to 

ensure it remains within current accident 

analysis assumptions. 

This proposed change does not impact the 

method of SG operation or primary or 

secondary coolant chemistry controls. In 

addition, this proposed change does not 

impact any other plant system or component. 

The change enhances SG inspection 

requirements. 

Therefore, this proposed change does not 

create the possibility of a new or different 

type of accident from any accident 

previously evaluated. 

Post-Modification ILRT Exception 

The proposed change would provide 

Callaway Plant with an exception from 

performing a post-modification containment 

Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 68187

integrated leak rate test following the 

replacement of the steam generators during 

Refuel 14. Providing an exception from 

performing a test does not involve a physical 

change to the plant nor does it change the 

operation of the plant. Thus it cannot 

introduce a new failure mode. Therefore 

adding a Technical Specification requirement 

that provides an exception from the steam 

generator replacement post-modification 

integrated leak rate testing requirement does 

not create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. The proposed changes do not involve a 

significant reduction in a margin of safety. 

Nuclear Steam Supply System Evaluations 

for Replacement Steam Generators 

The analyses and evaluations supporting 

the proposed RSG conditions reflect the 

reactor core safety limits. All acceptance 

criteria continue to be met. 

The analyses supporting the proposed RSG 

conditions reflect the proposed RTS and 

ESFAS Allowable Values. Setpoint 

calculations demonstrate that margin exists 

between these Allowable Values and the 

corresponding safety analysis limits used in 

the RSG analyses. The calculations are based 

on plant instrumentation and calibration/

functional test methods and include 

allowances for the RSG conditions. All 

analyses and evaluations supporting the 

proposed RSG core safety limits, decrease in 

maximum allowable power level for 3 

operable MSSVs per SG, the change in the 

manner in which the Reactor Coolant Flow—

Low Allowable Value is defined (while 

retaining the same numerical value), the 

change in the manner in which RCS average 

temperature is defined and the reduced 

upper limit for nominal T-avg at full power 

conditions in the Overtemperature DT and 

Overpower DT setpoint equations, and the 

changes to the pressurizer pressure and RCS 

average temperature limits in the DNB LCO 

[TS] 3.4.1 are acceptable. All acceptance 

criteria continue to be met. Therefore, the 

proposed changes do not involve a 

significant reduction in the margin of safety. 

Trip Time Delay Elimination 

This proposed change does not eliminate 

any RTS or ESFAS surveillances or alter the 

frequency of those surveillances as required 

by the TS. The SG Water Level Low—Low 

safety analysis limit of 0% span assumed in 

the analyses supporting the approval of the 

TTD design in Callaway Amendment 43 

dated April 14, 1989 is also used in the RSG 

analyses discussed above. None of the 

acceptance criteria for any accident analysis 

is changed for TTD elimination. 

There will be no effect on the manner in 

which safety limits or limiting safety system 

settings are determined nor will there be any 

effect on those plant systems necessary to 

assure the accomplishment of protection 

functions. The radiological dose consequence 

acceptance criteria will continue to be met. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

TSTF–449 Generic Licensing Change 

Package 

The SG tubes in pressurized water reactors 

are an integral part of the reactor coolant 

pressure boundary and, as such, are relied 

upon to maintain the primary system’s 

pressure and inventory. As part of the reactor 

coolant pressure boundary, the SG tubes are 

unique in that they are also relied upon as 

a heat transfer surface between the primary 

and secondary systems such that residual 

heat can be removed from the primary 

system. In addition, the SG tubes also isolate 

the radioactive fission products in the 

primary coolant from the secondary system. 

In summary, the safety function of a SG is 

maintained by ensuring the integrity of its 

tubes. 

Steam generator tube integrity is a function 

of the design, environment, and the physical 

condition of the tube. This proposed change 

to implement TSTF–449 does not, of itself, 

affect tube design or operating environment. 

The proposed change is expected to result in 

an improvement in the tube integrity by 

implementing the Steam Generator Program 

to manage SG tube inspection, assessment, 

repair (only under NRC-approved methods, 

none of which currently apply to the RSGs), 

and plugging. The requirements established 

by the Steam Generator Program are 

consistent with those in the applicable 

design codes and standards and are an 

improvement over the requirements in the 

existing TS.

For the above reasons, the margin of safety 

is not changed and overall plant safety will 

be enhanced by this proposed change. 

Post-Modification ILRT Exception 

The proposed change would provide 

Callaway Plant with an exception from 

performing a post-modification containment 

integrated leak rate test following the 

replacement of the steam generators during 

Refuel 14. The intent of post-modification 

integrated leak rate testing requirements is to 

assure the leak-tight integrity of the area 

affected by the modification. This intent will 

be satisfied by performing American Society 

of Mechanical Engineers code required 

inspections and tests. The acceptance 

criterion for American Society of Mechanical 

Engineers code system pressure testing for 

the base metal and welds is no leakage. In 

addition, the test pressure for the system 

pressure test will be several times that 

required during an integrated leak rate test. 

Since the leak-tight integrity of the primary 

containment boundary affected by the steam 

generator replacement will be assured, there 

is no change in the primary containment 

boundary’s ability to confine radioactive 

materials during an accident. Therefore, 

adding a Technical Specification requirement 

that provides an exception from the steam 

generator replacement post-modification 

integrated leak rate testing requirements 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: 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. 50–338 and 50–339, North 

Anna Power Station, Units No. 1 and 

No. 2, Louisa 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 U.S. 

Nuclear Regulatory Commission (NRC) 

staff review. 

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 North Anna 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, the 

location of the TS title and plant-specific title 

correlation, and the location of the unit staff 

qualification requirements. Therefore, this 

change has no bearing on the probability of 

an accident. Management organizational 

structure and safety and operational reviews 

have not changed and there is no change in 

the method of plant operation, operation 

review, or system design review. 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 

68188 Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 

Specifications. The proposed change only 

affects the designated titles of personnel, the 

location of the TS title and plant-specific title 

correlation, and the location of the unit staff 

qualification requirements. This change does 

not impact plant SSCs or plant operation. 

Management organizational structure and 

safety and operational reviews have not 

changed and 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. Therefore, 

the proposed change does not create the 

possibility of an accident of a different type 

than evaluated previously. 

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, the location of 

the TS title and plant-specific title 

correlation, and the location of the unit staff 

qualification requirements. This change does 

not impact plant design, plant operation or 

any safety margin. Therefore, the proposed 

change does not significantly reduce a 

margin of safety. 

This evaluation concludes that the 

proposed amendments to the North Anna 

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 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: October 

7, 2004. 

Description of amendment request:

The proposed change would revise 

Technical Specification (TS) 5.3, ‘‘Unit 

Staff Qualifications,’’ to reinstate the 

qualification requirements for the shift 

manager and control room supervisor 

positions that were inadvertently 

eliminated through Amendment No. 

150. Also, TS 5.3 would be revised to 

reference this amendment application 

for the use of the National Academy for 

Nuclear Training guideline, ACAD 00–

003, Revision 1, ‘‘Guidelines for Initial 

Training and Qualification of Licensed 

Operators.’’ Various other TSs would be 

revised to make corrections that were 

identified by the NRC staff in its letter 

dated January 28, 2004, and additional 

reviews performed by the licensee. 

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. 

Unit Staff Qualifications 

The proposed change is an administrative 

change to reinstate the qualification 

requirements for specific control room 

positions that were inadvertently eliminated 

through the issuance of Amendment No. 150 

and utilize Revision 1 to ACAD 00–003, 

‘‘Guidelines for Initial Training and 

Qualification of Licensed Operators.’’ The 

proposed change does not directly impact 

accidents previously evaluated. WCNOC’s 

[Wolf Creek Nuclear Operating Corporation’s] 

licensed operator training program is 

accredited by the National Academy for 

Nuclear Training and is based on a systems 

approach to training consistent with the 

requirements of 10 CFR 55. Although 

licensed operator qualifications and training 

may have an indirect impact on accidents 

previously evaluated, the NRC considered 

this impact during the rulemaking process, 

and by promulgation of the revised 10 CFR 

55 rule, concluded that this impact remains 

acceptable as long as the licensed operator 

training program is certified to be accredited 

and is based on a systems approach to 

training. 

Corrections 

The proposed change involves corrections 

to the Technical Specifications that are either 

associated with the issuance of the Improved 

Technical Specifications (Amendment No. 

123) or subsequent amendments. The 

changes are considered administrative 

changes and do not modify, add, delete, or 

relocate any technical requirements of the 

Technical Specifications. As such, 

administrative changes do not effect 

initiators of analyzed events or assumed 

mitigation of accident or transient events. 

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.

Unit Staff Qualifications 

The proposed change is an administrative 

change to reinstate the current requirements 

of specific control room positions and allow 

the use of Revision 1 of ACAD 00–003 for 

initial training and qualification of licensed 

operators. WCNOC’s licensed operator 

training program is accredited by the 

National Academy for Nuclear Training and 

is based on a systems approach to training 

consistent with the requirements of 10 CFR 

55. Although licensed operator qualifications 

and training may have an indirect impact on 

accidents previously evaluated, the NRC 

considered this impact during the 

rulemaking process, and by promulgation of 

the revised 10 CFR 55 rule, concluded that 

this impact remains acceptable as long as the 

licensed operator training program is 

certified to be accredited and is based on a 

systems approach to training. 

Corrections 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or changes in methods of governing normal 

plant operation. The proposed change will 

not impose any new or eliminate any old 

requirements. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. The proposed change does not involve 

a significant reduction in a margin of safety. 

Unit Staff Qualifications 

The proposed change is an administrative 

change to reinstate the current requirements 

of specific control room positions and allow 

the use of Revision 1 of ACAD 00–003 for 

initial training and qualification of licensed 

operators. As noted previously, WCNOC’s 

licensed operator training program is 

accredited and is based on a systems 

approach to training consistent with the 

requirements of 10 CFR 55. Licensed operator 

qualifications and training can have an 

indirect impact on the margin of safety. 

However, the NRC considered this impact 

during the rulemaking process, and by 

promulgation of the revised 10 CFR 55 rule, 

determined that this impact remains 

acceptable when licensees maintain a 

licensed operator training program that is 

accredited and based on a systems approach 

to training. 

Corrections 

The proposed change will not reduce a 

margin of safety because they have no effect 

on any safety analysis assumptions. The 

change is administrative in nature. 

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: Jay Silberg, Esq., 

Shaw, Pittman, Potts and Trowbridge, 

2300 N Street, NW., Washington, DC 

20037. 

Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 68189

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. 

FPL Energy Seabrook, LLC, Docket No. 

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

Rockingham County, New Hampshire 

Date of amendment request: October 

22, 2004. 

Description of amendment request:

The proposed amendment would revise 

the allowed outage times of Technical 

Specification 3.3.3.6, ‘‘Accident 

Monitoring Instrumentation,’’ to be 

consistent with the completion times in 

the related specification in NUREG–

1431, Revision 3, ‘‘Standard Technical 

Specifications Westinghouse Plants.’’

Date of publication of individual 

notice in Federal Register: November 

2, 2004 (69 FR 63560). 

Expiration date of individual notice:

December 2, 2004 (public comments) 

and January 3, 2005 (hearing requests).

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

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 email 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:

December 23, 2003, as supplemented by 

letter dated June 16, 2004. 

Brief description of amendment: The 

amendment revised Section 4.5.D of the 

Technical Specifications to specify 

testing the main steam isolation valves 

at a pressure lower than Pa, the 

calculated peak containment internal 

pressure related to the design-basis lossof-coolant accident. 

Date of Issuance: November 2, 2004. 

Effective date: November 2, 2004 and 

shall be implemented within 30 days of 

issuance 

Amendment No.: 250. 

Facility Operating License No. DPR–

16: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: February 17, 2004 (69 FR 

7518). 

The June 16, 2004, letter 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 2, 2004. 

No significant hazards consideration 

comments received: No. 

Duke Energy Corporation, Docket Nos. 

50–269, 50–270, and 50–287, Oconee 

Nuclear Station, Units 1, 2, and 3, 

Oconee County, South Carolina 

Date of application of amendments:

January 15, 2004, as supplemented by 

letter dated March 15, 2004. 

Brief description of amendments: The 

amendments revise the Technical 

Specifications associated with the 

control rod drive trip devices. The 

amendments are needed to support 

implementation of the reactor trip 

breaker replacement. 

Date of Issuance: November 2, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: 341, 343, 342. 

Renewed Facility Operating License 

Nos. DPR–38, DPR–47, and DPR–55:

Amendments revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: April 13, 2004 (69 FR 19566). 

The supplement dated March 15, 2004, 

provided clarifying information that did 

not change the scope of the January 15, 

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 2, 

2004. 

No significant hazards consideration 

comments received: No.

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 application for amendments:

March 22, 2004, as supplemented July 

23 and October 11, 2004. 

Brief description of amendments: The 

amendments modified Technical 

Specification (TS) requirements to adopt 

the provisions of Industry/TS Task 

68190 Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 

Force (TSTF) change TSTF–359, 

‘‘Increased Flexibility in Mode 

Restraints.’’

Date of issuance: November 4, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 263 and 144. 

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised 

the TSs. 

Date of initial notice in Federal 

Register: August 31, 2004 (69 FR 

53108). The supplemental letters dated 

July 23 and October 11, 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 determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 4, 

2004. 

No significant hazards consideration 

comments received: No. 

Nine Mile Point Nuclear Station, LLC, 

Docket Nos. 50–220, and 50–410, Nine 

Mile Point Nuclear Station, Unit Nos. 1 

and 2, Oswego County, New York 

Date of application for amendments:

January 8, 2004 (2 letters), as 

supplemented by letter dated June 17, 

2004. 

Brief description of amendments: The 

amendments approve implementation of 

the Boiling Water Reactor Vessel and 

Internals Project Reactor Pressure Vessel 

Integrated Surveillance Program as the 

basis for demonstrating the units’

compliance with the requirements of 

appendix H to Title 10 of the Code of 

Federal Regulations. Specifically, the 

amendments approved the wording 

proposed by the licensee to update the 

units’ Updated Safety Analysis Reports. 

In addition, the Unit 1 amendment also 

revised the Technical Specifications to 

delete any reference to plant-specific 

surveillance requirements. 

Date of issuance: November 8, 2004. 

Effective date: As of the date of 

issuance. Integrated Surveillance 

Program shall be implemented within 

90 days of issuance. The units’ Final 

Safety Analysis Report (Updated) shall 

be updated in accordance with 10 CFR 

50.71(e). 

Amendment Nos.: 184 and 114. 

Facility Operating License Nos. DPR–

63 and NPF–69: Amendments revise the 

Technical Specifications (for Unit 1), 

the operating license (for Unit 2), and 

approve revision of licensing basis for 

both units. 

Date of initial notice in Federal 

Register: February 17, 2004 (69 FR 

7524). The June 17, 2004, letter 

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 the 

amendments is contained in two Safety 

Evaluations, both dated November 8, 

2004. 

No significant hazards consideration 

comments received: No.

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant, Wright County, 

Minnesota 

Date of application for amendment:

December 23, 2003, as supplemented 

June 21, 2004. 

Brief description of amendment: The 

amendment changes Technical 

Specification (TS) Limiting Condition 

for Operation (LCO) Tables 3.2.1 and 

3.2.4 to (1) eliminate the reactor head 

cooling containment isolation function 

from the TSs, (2) correct and clarify the 

description of the number of instrument 

channels per trip system as defined in 

the TSs, and (3) revise an existing LCO 

for radiation monitors used to isolate 

reactor building ventilation and initiate 

the standby gas treatment system. 

Date of issuance: November 2, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 140. 

Facility Operating License No. DPR–

22. Amendment revised the TSs. 

Date of initial notice in Federal 

Register: March 30, 2004 (69 FR 

16621). 

The supplemental letter contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination and did not 

expand the scope of the original Federal 

Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated November 2, 

2004. 

No significant hazards consideration 

comments received: No. 

Pacific Gas and Electric Company, 

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

Canyon Nuclear Power Plant, Unit Nos. 

1 and 2, San Luis Obispo County, 

California 

Date of application for amendments:

March 18, 2004, and its supplements 

dated August 18 and 20, and September 

17, 2004. 

Brief description of amendments: The 

amendments authorize revisions to the 

Final Safety Analysis Report (FSAR) 

Update to incorporate the NRC approval 

of a permanently revised steam 

generator voltage-based repair criteria 

probability of detection (POD) method. 

The revised POD method is referred to 

as the probability of prior cycle 

detection method. In addition, a 

reporting requirement is added to the 

DCPP Technical Specifications as TS 

5.6.10.i. 

Date of issuance: October 28, 2004. 

Effective date: October 28, 2004, and 

shall be implemented within 30 days of 

the date of issuance. The 

implementation of the amendment 

includes the incorporation into the 

FSAR Update the changes discussed 

above, as described in the licensee’s 

application dated March 18, 2004, and 

its supplements dated August 18 and 

20, and September 17, 2004, and 

evaluated in the staff’s Safety Evaluation 

attached to the amendments. 

Amendment Nos.: Unit 1–177; Unit 

2–179. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the FSAR Update and the 

Technical Specifications. 

Date of initial notice in Federal 

Register: June 22, 2004 (69 FR 34704). 

The August 18 and 20, and September 

17, 2004, supplemental letters provided 

additional clarifying information, did 

not expand the scope of the application 

as originally noticed, and did not 

change the staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 28, 

2004. 

No significant hazards consideration 

comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50–

387 and 50–388, Susquehanna Steam 

Electric Station, Units 1 and 2 (SSES 1 

and 2), Luzerne County, Pennsylvania 

Date of application for amendments:

December 22, 2003, as supplemented by 

letters dated June 18, July 15, and 

September 8, 2004. 

Brief description of amendments: The 

amendment added TS 3.3.1.3, 

‘‘Oscillation Power Range Monitor 

(OPRM) Instrumentation,’’ and changed 

TS 3.4.1, ‘‘Recirculation Loops 

Operating,’’ and TS 5.6.5, ‘‘Core 

Operating Limits Report,’’ to remove 

specifications and information related to 

current stability specifications which 

will no longer be needed with the 

operation of the OPRM system. 

Date of issuance: November 9, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment Nos.: 217 and 192. 

Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 68191

Facility Operating License Nos. NPF–

14 and NPF–22: The amendment 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: January 20, 2004 (69 FR 

2745). The supplements dated June 18, 

July 15, and September 8, 2004, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 9, 

2004. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of application for amendment:

March 31, 2003, as supplemented by 

letter dated July 30, 2004. 

Brief description of amendment: The 

amendment revised the reactor pressure 

vessel pressure-temperature limits and 

extends their validity to 32 effective full 

power years. 

Date of issuance: November 1, 2004. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 157. 

Facility Operating License No. NPF–

57: This amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: June 8, 2004 (69 FR 32076). 

The July 30, 2004 letter provided 

clarifying information that did not 

change the initial proposed no 

significant hazards consideration 

determination or expand the application 

beyond the scope of the original Federal 

Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated November 1, 

2004. 

No significant hazards consideration 

comments received: No.

Notice of Issuance of Amendments to 

Facility Operating Licenses and Final 

Determination of No Significant 

Hazards Consideration and 

Opportunity for a Hearing (Exigent 

Public Announcement or Emergency 

Circumstances) 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application for the 

amendment complies with the 

standards and requirements of the 

Atomic Energy Act of 1954, as amended 

(the Act), and the Commission’s rules 

and regulations. The Commission has 

made appropriate findings as required 

by the Act and the Commission’s rules 

and regulations in 10 CFR Chapter I, 

which are set forth in the license 

amendment. 

Because of exigent or emergency 

circumstances associated with the date 

the amendment was needed, there was 

not time for the Commission to publish, 

for public comment before issuance, its 

usual Notice of Consideration of 

Issuance of Amendment, Proposed No 

Significant Hazards Consideration 

Determination, and Opportunity for a 

Hearing. 

For exigent circumstances, the 

Commission has either issued a Federal 

Register notice providing opportunity 

for public comment or has used local 

media to provide notice to the public in 

the area surrounding a licensee’s facility 

of the licensee’s application and of the 

Commission’s proposed determination 

of no significant hazards consideration. 

The Commission has provided a 

reasonable opportunity for the public to 

comment, using its best efforts to make 

available to the public means of 

communication for the public to 

respond quickly, and in the case of 

telephone comments, the comments 

have been recorded or transcribed as 

appropriate and the licensee has been 

informed of the public comments. 

In circumstances where failure to act 

in a timely way would have resulted, for 

example, in derating or shutdown of a 

nuclear power plant or in prevention of 

either resumption of operation or of 

increase in power output up to the 

plant’s licensed power level, the 

Commission may not have had an 

opportunity to provide for public 

comment on its no significant hazards 

consideration determination. In such 

case, the license amendment has been 

issued without opportunity for 

comment. If there has been some time 

for public comment but less than 30 

days, the Commission may provide an 

opportunity for public comment. If 

comments have been requested, it is so 

stated. In either event, the State has 

been consulted by telephone whenever 

possible. 

Under its regulations, the Commission 

may issue and make an amendment 

immediately effective, notwithstanding 

the pendency before it of a request for 

a hearing from any person, in advance 

of the holding and completion of any 

required hearing, where it has 

determined that no significant hazards 

consideration is involved. 

The Commission has applied the 

standards of 10 CFR 50.92 and has made 

a final determination that the 

amendment involves no significant 

hazards consideration. The basis for this 

determination is contained in the 

documents related to this action. 

Accordingly, the amendments have 

been issued and made effective as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the application for 

amendment, (2) the amendment to 

Facility Operating License, and (3) the 

Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment, as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the 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.) 

The Commission is also offering an 

opportunity for a hearing with respect to 

the issuance of the amendment. Within 

60 days after the date of publication of 

this notice, the licensee may file a 

request for a hearing with respect to 

issuance of the amendment to the 

subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

68192 Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 

1To the extent that the applications contain 

attachments and supporting documents that are not 

publicly available because they are asserted to 

contain safeguards or proprietary information, 

petitioners desiring access to this information 

should contact the applicant or applicant’s counsel 

and discuss the need for a protective order.

intervene. Requests for a hearing and a 

petition for leave to intervene shall be 

filed in accordance with the 

Commission’s ‘‘Rules of Practice for 

Domestic Licensing Proceedings’’ in 10 

CFR Part 2. Interested persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland, 

and electronically on the Internet at the 

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

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

are problems in accessing the document, 

contact the PDR Reference staff at 1–

800–397–4209, 301–415–4737, or by email to pdr@nrc.gov. (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 by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the petitioner/

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner intends to 

rely in proving the contention at the 

hearing. The petitioner must also 

provide references to those specific 

sources and documents of which the 

petitioner is aware and on which the 

petitioner intends to rely to establish 

those facts or expert opinion. The 

petition must include sufficient 

information to show that a genuine 

dispute exists with the applicant on a 

material issue of law or fact.1

Contentions shall be limited to matters 

within the scope of the amendment 

under consideration. The contention 

must be one which, if proven, would 

entitle the petitioner to relief. A 

petitioner/requestor who fails to satisfy 

these requirements with respect to at 

least one contention will not be 

permitted to participate as a party.

Each contention shall be given a 

separate numeric or alpha designation 

within one of the following groups: 

1. Technical—primarily concerns/

issues relating to technical and/or 

health and safety matters discussed or 

referenced in the applications. 

2. Environmental—primarily 

concerns/issues relating to matters 

discussed or referenced in the 

environmental analysis for the 

applications. 

3. Miscellaneous—does not fall into 

one of the categories outlined above. 

As specified in 10 CFR 2.309, if two 

or more petitioners/requestors seek to 

co-sponsor a contention, the petitioners/

requestors shall jointly designate a 

representative who shall have the 

authority to act for the petitioners/

requestors with respect to that 

contention. If a petitioner/requestor 

seeks to adopt the contention of another 

sponsoring petitioner/requestor, the 

petitioner/requestor who seeks to adopt 

the contention must either agree that the 

sponsoring petitioner/requestor shall act 

as the representative with respect to that 

contention, or jointly designate with the 

sponsoring petitioner/requestor a 

representative who shall have the 

authority to act for the petitioners/

requestors with respect to that 

contention.

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. Since the Commission has 

made a final determination that the 

amendment involves no significant 

hazards consideration, if a hearing is 

requested, it will not stay the 

effectiveness of the amendment. Any 

hearing held would take place while the 

amendment is in effect. 

A request for a hearing or a petition 

for leave to intervene must be filed by: 

(1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, Attention: Rulemaking and 

Adjudications Staff; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) E-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

hearingdocket@nrc.gov; or (4) facsimile 

transmission addressed to the Office of 

the Secretary, U.S. Nuclear Regulatory 

Commission, Washington, DC, 

Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

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

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer or 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(i)–(viii). 

STP Nuclear Operating Company, 

Docket No. 50–499, South Texas Project, 

Unit 2, Matagorda County, Texas 

Date of amendment request:

September 30, 2004. 

Description of amendment request:

The amendment changes Technical 

Specification 4.4.4.2 to expand the 

range of conditions under which 

quarterly testing of block valves for the 

pressurizer power operated relief valves 

would be unnecessary. 

Date of issuance: October 21, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 153. 

Federal Register / Vol. 69, No. 225 / Tuesday, November 23, 2004 / Notices 68193

Facility Operating License No. NPF–

80: Amendment revises the Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): Yes. October 6, 

2004 (69 FR 59969). 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 December 6, 2004, 

but indicated that if the Commission 

makes a final NSHC determination, any 

such hearing would take place after 

issuance of the amendment. 

The Commission’s related evaluation 

of the amendment, finding of exigent 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated October 21, 

2004. 

Attorney for licensee: Mr. John E. 

Matthews, Morgan, Lewis & Bokius, 

LLP, 1111 Pennsylvania Avenue, NW., 

Washington, DC 20004. 

NRC Section Chief: Michael K. Webb, 

Acting.

Nuclear Management Company, LLC, 

Docket No. 50–255, Palisades Plant, Van 

Buren County, Michigan 

Date of amendment request:

November 2, 2004. 

Description of amendment request:

The amendment revises Technical 

Specification Limiting Condition for 

Operation 3.4.3, ‘‘Primary Coolant 

System (PCS) Pressure and Temperature 

(P/T) Limits’’ to add restrictions to the 

cooldown rate limits. This amendment 

supports plant restart following repairs 

of two reactor vessel closure head 

control rod drive nozzle penetrations at 

the Palisades Nuclear Power Plant. 

Date of issuance: November 8, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

immediately. 

Amendment No.: 218. 

Facility Operating License No. DPR–

20: Amendment revises the Technical 

Specification. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC):

The Commission’s related evaluation 

of the amendment, finding of emergency 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated November 

8, 2004. 

Attorney for licensee: Jonathan Rogoff, 

Esquire, Vice President, Counsel & 

Secretary, Nuclear Management 

Company, LLC, 700 First Street, 

Hudson, WI 54016. 

NRC Section Chief: L. Raghavan. 

Tennessee Valley Authority, Docket No. 

50–327, Sequoyah Nuclear Plant, Unit 1, 

Hamilton County, Tennessee 

Date of amendment request:

November 4, 2004. 

Description of amendment request:

The proposed amendment extended the 

implementation period for License 

Amendment 294 to May 15, 2005. 

Date of issuance: November 9, 2004. 

Effective date: As of date of issuance, 

to be implemented by May 15, 2005. 

Amendment No.: 297. 

Facility Operating License No. DPR–

77: Amendment revises the 

implementation date for License 

Amendment No. 294. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC):

The Commission’s related evaluation 

of the amendment, finding of emergency 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated November 

9, 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 15th 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. 04–25664 Filed 11–22–04; 8:45 am] 

BILLING CODE 7590–01–P

OFFICE OF PERSONNEL 

MANAGEMENT 

Excepted Service

AGENCY: Office of Personnel 

Management.

ACTION: Notice.

SUMMARY: This gives notice of OPM 

decisions granting authority to make 

appointments under Schedules A, B and 

C in the excepted service as required by 

5 CFR 6.6 and 213.103.

FOR FURTHER INFORMATION CONTACT:

Hughes Turner, Deputy Associate 

Director, Center for Leadership and 

Executive Resources Policy, Division for 

Strategic Human Resources Policy, 202–

606–1811.

SUPPLEMENTARY INFORMATION: Appearing 

in the listing below is one Schedule A 

appointment, no Schedule B 

appointments, and Schedule C 

appointments established between 

October 1, 2004 and October 31, 2004. 

Future notices will be published on the 

fourth Tuesday of each month, or as 

soon as possible thereafter. A 

consolidated listing of all authorities as 

of June 30 is published each year. 

Schedule A 

Department of Homeland Security 

213.3111 

Up to 15 Senior Level and General 

Schedule (or equivalent) positions 

within the Homeland Security Labor 

Relations Board and the Homeland 

Security Mandatory Removal Panel. 

Effective October 15, 2004. 

Schedule B 

No Schedule B appointments for 

October 2004. 

Schedule C 

The following Schedule C 

appointments were approved for 

October 2004: 

Section 213.3303 Executive Office of 

the President, Office of National Drug 

Control Policy 

QQGS00083 Intergovernmental 

Affairs Liaison to the Chief of Staff. 

Effective October 19, 2004. 

QQGS00086 Legislative Assistant to 

the Associate Director, Legislative 

Affairs. Effective October 19, 2004. 

Section 213.3304 Department of State 

DSGS60797 Legislative Management 

Officer to the Assistant Secretary for 

Legislative and Intergovernmental 

Affairs. Effective October 14, 2004. 

DSGS60798 Legislative Management 

Officer to the Assistant Secretary for 

Legislative and Intergovernmental 

Affairs. Effective October 14, 2004. 

DSGS60799 Foreign Affairs Officer to 

the Under Secretary for Global Affairs. 

Effective October 14, 2004. 

DSGS60800 Staff Assistant to the 

Senior Advisor to the Secretary and 

White House Liaison. Effective October 

28, 2004. 

Section 213.3304 Department of 

Treasury 

DYGS00434 Special Assistant to the 

Deputy Chief of Staff. Effective October 

25, 2004. 

Section 213.3306 Department of 

Defense 

DDGS16831 Research Assistant to the 

Deputy Assistant Secretary of Defense 

(Strategic Communications Planning). 

Effective October 1, 2004. 

DDGS16842 Staff Assistant to the 

Deputy Assistant Secretary of Defense 

(Special Operations and Combating 

Terrorism). Effective October 20, 2004.