Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62467

matters related to the conduct of 

Committee activities and matters and 

specific issues that were not completed 

during previous meetings, as time and 

availability of information permit. 

Procedures for the conduct of and 

participation in ACRS meetings were 

published in the Federal Register on 

October 5, 2004 (69 FR 59620). In 

accordance with those procedures, oral 

or written views may be presented by 

members of the public, including 

representatives of the nuclear industry. 

Electronic recordings will be permitted 

only during the open portions of the 

meeting. Persons desiring to make oral 

statements should notify the Cognizant 

ACRS staff named below five days 

before the meeting, if possible, so that 

appropriate arrangements can be made 

to allow necessary time during the 

meeting for such statements. Use of still, 

motion picture, and television cameras 

during the meeting may be limited to 

selected portions of the meeting as 

determined by the Chairman. 

Information regarding the time to be set 

aside for this purpose may be obtained 

by contacting the Cognizant ACRS staff 

prior to the meeting. In view of the 

possibility that the schedule for ACRS 

meetings may be adjusted by the 

Chairman as necessary to facilitate the 

conduct of the meeting, persons 

planning to attend should check with 

the Cognizant ACRS staff if such 

rescheduling would result in major 

inconvenience. 

In accordance with subsection 10(d) 

Pub. L. 92–463, I have determined that 

it is necessary to close portions of this 

meeting noted above to discuss and 

protect information classified as 

national security information as well as 

safeguard information pursuant to 5 

U.S.C. 552b(c)(1) and (3). 

Further information regarding topics 

to be discussed, whether the meeting 

has been canceled or rescheduled, as 

well as the Chairman’s ruling on 

requests for the opportunity to present 

oral statements and the time allotted 

therefor can be obtained by contacting 

Mr. Sam Duraiswamy, Cognizant ACRS 

staff (301–415–7364), between 7:30 a.m. 

and 4:15 p.m., e.t. 

ACRS meeting agenda, meeting 

transcripts, and letter reports are 

available through the NRC Public 

Document Room at pdr@nrc.gov, or by 

calling the PDR at 1–800–397–4209, or 

from the Publicly Available Records 

System (PARS) component of NRC’s 

document system (ADAMS) which is 

accessible from the NRC Web site at 

http://www.nrc.gov/reading-rm/

adams.html or http://www.nrc.gov/

reading-rm/doc-collections/ (ACRS & 

ACNW Mtg schedules/agendas). 

Videoteleconferencing service is 

available for observing open sessions of 

ACRS meetings. Those wishing to use 

this service for observing ACRS 

meetings should contact Mr. Theron 

Brown, ACRS Audio Visual Technician 

(301–415–8066), between 7:30 a.m. and 

3:45 p.m., e.t., at least 10 days before the 

meeting to ensure the availability of this 

service. Individuals or organizations 

requesting this service will be 

responsible for telephone line charges 

and for providing the equipment and 

facilities that they use to establish the 

videoteleconferencing link. The 

availability of videoteleconferencing 

services is not guaranteed.

Dated: October 20, 2004. 

Andrew L. Bates, 

Advisory Committee Management Officer.

[FR Doc. 04–23903 Filed 10–25–04; 8:45 am] 

BILLING CODE 7590–01–P

NUCLEAR REGULATORY 

COMMISSION 

Sunshine Federal Register Notice

DATES: Weeks of October 25, November 

1, 8, 15, 22, 29, 2004.

PLACE: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland.

STATUS: Public and Closed.

MATTERS TO BE CONSIDERED: 

Week of October 25, 2004

There are no meetings scheduled for 

the week of October 25, 2004. 

Week of November 1, 2004—Tentative 

There are no meetings scheduled for 

the week of November 1, 2004. 

Week of November 8, 2004—Tentative 

Monday, November 8, 2004

9 a.m. Briefing on Plant Aging and 

Material Degradation Issues—Part 

One (Public Meeting) (Contact: 

Steve Koenick, 301–415–1239) 

1:30 p.m. Briefing on Plant Aging and 

Material Degradation Issues—Part 

Two (Public Meeting) (Contact: 

Steve Koenick, 301–415–1239)

This meeting (both parts) will be 

webcast live at the Web address—

http://www.nrc.gov.

Week of November 15, 2004—Tentative 

Tuesday, November 16, 2004

1:30 p.m. Briefing on Threat 

Environment Assessment (Closed—

Ex. 1) (New time) 

Thursday, November 18, 2004

1:30 p.m. Discussion of Security Issues 

(Closed—Ex. 1) (New date and time) 

Week of November 22, 2004—Tentative 

There are no meetings scheduled for 

the week of November 22, 2004. 

Week of November 29, 2004—Tentative 

There are no meetings scheduled for 

the week of November 29, 2004. 

* The schedule for Commission 

meetings is subject to change on short 

notice. To verify the status of meetings 

call (recording)—(301) 415–1292. 

Contact person for more information: 

Dave Gamberoni, (301) 415–1651. 

‘‘Briefing on Reactor Safety and 

Licensing Activities (Public Meeting),’’

originally scheduled for 9:30 a.m. on 

Tuesday, November 9, 2004, is being 

rescheduled for a later date. 

The NRC Commission Meeting 

Schedule can be found on the Internet 

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

policy-making/schedule.html.

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g.

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

August Spector, at 301–415–7080, TDD: 

301–415–2100, or by e-mail at 

aks@nrc.gov. Determinations on 

requests for reasonable accommodation 

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

This notice is distributed by mail to 

several hundred subscribers; if you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301–415–1969). 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to dkw@nrc.gov.

Dated: October 21, 2004. 

Dave Gameroni, 

Office of the Secretary.

[FR Doc. 04–24010 Filed 10–22–04; 10:12 

am] 

BILLING CODE 7590–01–M

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a.(2) of the 

Atomic Energy Act of 1954, as amended 

62468 Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 

(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 1, 

2004 through October 14, 2004. The last 

biweekly notice was published on 

October 12, 2004 (69 FR 60677). 

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 01F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. The filing of 

requests for a hearing and petitions for 

leave to intervene is discussed below. 

Within 60 days after the date of 

publication of this notice, the licensee 

may file a request for a hearing with 

respect to issuance of the amendment to 

the subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

intervene. Requests for a hearing and a 

petition for leave to intervene shall be 

filed in accordance with the 

Commission’s ‘‘Rules of Practice for 

Domestic Licensing Proceedings’’ in 10 

CFR part 2. Interested persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order.

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/

requestor to relief. A petitioner/

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62469

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. 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. 50–461, Clinton Power Station, Unit 

1, DeWitt County, Illinois 

Date of amendment request: April 30, 

2004.

Description of amendment request:

The proposed change allows entry into 

a mode or other specified condition in 

the applicability of a technical 

specification (TS), while in a condition 

statement and the associated required 

actions of the TS, provided the licensee 

performs a risk assessment and manages 

risk consistent with the program in 

place for complying with the 

requirements of title 10 of the Code of 

Federal Regulations (10 CFR), part 50, 

Section 50.65(a)(4). Limiting Condition 

for Operation (LCO) 3.0.4 exceptions in 

individual TS would be eliminated, 

several notes or specific exceptions are 

revised to reflect the related changes to 

LCO 3.0.4, and Surveillance 

Requirement (SR) 3.0.4 is revised to 

reflect the LCO 3.0.4 allowance. 

This change was proposed by the 

industry’s Technical Specification Task 

Force (TSTF) and is designated TSTF–

359. The Nuclear Regulatory 

Commission (NRC) staff issued a notice 

of opportunity for comment in the 

Federal Register on August 2, 2002 (67 

FR 50475), on possible amendments 

concerning TSTF–359, including a 

model safety evaluation and model no 

significant hazards consideration 

(NSHC) determination, using the 

consolidated line item improvement 

process. The NRC staff subsequently 

issued a notice of availability of the 

models for referencing in license 

amendment applications in the Federal 

Register on April 4, 2003 (68 FR 16579). 

The licensee affirmed the applicability 

of the following NSHC determination in 

its application dated April 30, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of NSHC is 

presented below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change allows entry into a 

mode or other specified condition in the 

applicability of a TS, while in a TS condition 

statement and the associated required actions 

of the TS. Being in a TS condition and the 

associated required actions is not an initiator 

of any accident previously evaluated. 

Therefore, the probability of an accident 

previously evaluated is not significantly 

increased. The consequences of an accident 

while relying on required actions as allowed 

by proposed LCO 3.0.4, are no different than 

the consequences of an accident while 

entering and relying on the required actions 

while starting in a condition of applicability 

of the TS. Therefore, the consequences of an 

accident previously evaluated are not 

significantly affected by this change. The 

addition of a requirement to assess and 

manage the risk introduced by this change 

will further minimize possible concerns. 

Therefore, this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Previously 

Evaluatedty. 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed). 

Entering into a mode or other specified 

condition in the applicability of a TS, while 

in a TS condition statement and the 

associated required actions of the TS, will 

not introduce new failure modes or effects 

and will not, in the absence of other 

unrelated failures, lead to an accident whose 

consequences exceed the consequences of 

accidents previously evaluated. The addition 

of a requirement to assess and manage the 

risk introduced by this change will further 

minimize possible concerns. Thus, this 

change does not create the possibility of a 

new or different kind of accident from an 

accident previously evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in a Margin 

of Safety 

The proposed change allows entry into a 

mode or other specified condition in the 

applicability of a TS, while in a TS condition 

statement and the associated required actions 

of the TS. The TS allow operation of the 

plant without the full complement of 

equipment through the conditions for not 

meeting the TS LCO. The risk associated with 

this allowance is managed by the imposition 

of required actions that must be performed 

within the prescribed completion times. The 

net effect of being in a TS condition on the 

62470 Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 

margin of safety is not considered significant. 

The proposed change does not alter the 

required actions or completion times of the 

TS. The proposed change allows TS 

conditions to be entered, and the associated 

required actions and completion times to be 

used in new circumstances. This use is 

predicated upon the licensee’s performance 

of a risk assessment and the management of 

plant risk. The change also eliminates current 

allowances for utilizing required actions and 

completion times in similar circumstances, 

without assessing and managing risk. The net 

change to the margin of safety is 

insignificant. Therefore, this change does not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration.

Attorney for licensee: Mr. Thomas S. 

O’Neill, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60666. 

NRC Section Chief: Gene Y. Suh. 

Calvert Cliffs Nuclear Power Plant, Inc., 

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

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of amendment request:

December 9, 2003. 

Description of amendment request:

The proposed amendment would revise 

Technical Specification 3.7.1, ‘‘Main 

Steam Safety Valves (MSSVs),’’ to 

increase the maximum allowable lift 

setting on two MSSVs on each unit. In 

addition, the proposed amendment 

would increase the completion time for 

reducing the Power Level-High Trip 

setpoint. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Would not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

This license amendment request proposes 

to increase the upper range of the relief 

setting of the first two Main Steam Safety 

Valves (MSSVs) by 10 psi [pounds per 

square-inch]. The MSSVs are not accident 

initiators. They are credited with relieving 

secondary system pressure and act as a heat 

sink for the Reactor Coolant System (RCS) 

when the preferred heat sink is not available. 

Increasing the upper end of the setpoint for 

the first two MSSVs to lift does not affect the 

steam relieving capacity of the total or any 

combination of MSSVs that lift. This 

proposed amendment does not install any 

new components or change the physical 

characteristics of the MSSVs. Therefore, the 

change does not involve a significant 

increase in the probability of an evaluated 

accident. 

The Updated Final Safety Analysis Report 

Chapter 14 safety analyses were reviewed 

considering the change to the upper end of 

the lift settings range of the first two MSSVs. 

The analyses show that increasing the upper 

end of the lift setting range does not exceed 

the pressure limits of the reactor coolant or 

main steam systems, nor the radiological 

consequences anticipated by the safety 

analyses. Therefore, the change will not 

involve a significant increase in the 

consequences of an evaluated accident. 

This proposed amendment will also 

increase the Technical Specification 

Completion Time to reset the Power LevelHigh Trip from 12 hours to 36 hours. The 

purpose of the Power Level-High Trip is to 

trip the reactor if reactor power exceeds a set 

value, and is required by Technical 

Specifications to be reset according to the 

number of MSSVs remaining operable. The 

trip is not an accident initiator but is a signal 

that responds to an accident condition. 

Therefore, the change does not involve a 

significant increase in the probability of an 

evaluated accident. 

Reducing the setpoint of the Power LevelHigh Trip within the time allotted by 

Technical Specifications provides additional 

assurance that the MSSVs will be able to 

perform their design function by keeping the 

reactor power within the ability of the 

MSSVs to relieve steam volume. There is low 

probability of a transient that could result in 

steam generator overpressure during the 

proposed 36 hours to reset the Power LevelHigh Trip. Therefore, this change does not 

involve a significant increase in the 

consequences of an evaluated accident. 

Therefore, this proposed license 

amendment does not significantly increase 

the probability or consequences of an 

accident previously evaluated. 

2. Would not create the possibility of a new 

or different [kind] of accident from any 

accident previously evaluated. 

The proposed amendment will increase the 

upper end of the lift pressure for the first two 

MSSVs and increase the Technical 

Specification Completion Time to reset the 

Power Level-High Trip setpoint. 

The proposed amendment does not involve 

a physical alteration of the plant or change 

the plant configuration. It does not require 

any new or unusual operator actions. The 

amendment does not alter the way any 

structure, system, or component functions 

and does not alter the manner in which the 

plant is operated. It does not introduce any 

new failure modes. 

Therefore, this proposed license 

amendment does not create the possibility of 

a new or different [kind] of accident from any 

accident previously evaluated. 

3. Would not involve a significant 

reduction in [a] margin of safety. 

The margin of safety in this case is that the 

MSSVs release sufficient steam to relieve 

pressure in the secondary system and to act 

as a heat sink to prevent over-pressurization 

of the RCS when the preferred heat sink is 

not available. Increasing the upper end of the 

setpoint for the first two MSSVs to lift does 

not affect the steam relieving capacity of the 

total or any combination of MSSVs that lift. 

Potential delay in the opening of the first two 

MSSVs does not result in exceeding the 

pressure limits of the reactor coolant or main 

steam systems.

Reducing the Power Level-High Trip 

setpoint within the specified time limit 

provides additional assurance that the 

MSSVs will be able to perform their design 

function by keeping the reactor power within 

the ability of the MSSVs to relieve steam 

volume. A completion time of 36 hours to 

lower the Power Level-High Trip setpoint is 

based on a reasonable time to correct the 

MSSV inoperability, operating experience in 

resetting all channels of a protective 

function, and on the low probability of the 

occurrence of a transient that could result in 

steam generator overpressure during this 

period. 

Therefore, this proposed license 

amendment does not involve a significant 

reduction in [a] margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: James M. Petro, 

Jr., Esquire, Counsel, Constellation 

Energy Group, Inc., 750 East Pratt Street, 

5th floor, Baltimore, MD 21202. 

NRC Section Chief: Richard J. Laufer. 

Carolina Power & Light Company, 

Docket Nos. 50–325 and 50–324, 

Brunswick Steam Electric Plant, Units 1 

and 2, Brunswick County, North 

Carolina 

Date of amendments request: August 

16, 2004. 

Description of amendments request:

The proposed change adds topical 

report NEDE–32906P–A, ‘‘TRACG 

Application for Anticipated Operational 

Occurrences (AOO) Transient 

Analyses,’’ to the documents listed in 

Technical Specification (TS) 5.6.5 

describing the approved methodologies 

used to determine the core operating 

limits. Unit 2 will be unable to resume 

power operation following Refueling 

Outage 16 without NRC approval for 

inclusion of the TRACG methodology in 

TS 5.6.5.b. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62471

Response: No. 

The proposed change to TS 5.6.5.b will add 

General Electric Nuclear Energy topical 

report NEDE–32906P–A, ‘‘TRACG 

Application for Anticipated Operational 

Occurrences (AOO) Transient Analyses,’’ to 

the list of documents describing approved 

methodologies for determining core operating 

limits. NRC review and acceptance of the 

TRACG methodology is documented in an 

October 22, 2001, letter and associated safety 

evaluation issued to General Electric Nuclear 

Energy (i.e., refer to ADAMS Accession 

Numbers ML012740390 and ML012740161). 

Analyzed events are assumed to be initiated 

by the failure of plant structures, systems, or 

components. The core operating limits, 

which are developed using the topical report 

being added, ensure that the integrity of the 

fuel will be maintained during normal 

operations and that design requirements will 

continue to be met. The proposed change 

does not involve physical changes to any 

plant structure, system, or component. 

Therefore, the probability of occurrence for a 

previously analyzed accident is not 

significantly increased. 

The consequences of a previously analyzed 

accident are dependent on the initial 

conditions assumed for the analysis, the 

behavior of the fuel during the analyzed 

accident, the availability and successful 

functioning of the equipment assumed to 

operate in response to the analyzed event, 

and the setpoints at which these actions are 

initiated. Use of the analytical methodologies 

described in the topical report being added 

to TS 5.6.5.b will ensure that applicable 

design and safety analyses acceptance criteria 

are met. Use of these NRC-approved 

methodologies does not affect the 

performance of any equipment used to 

mitigate the consequences of an analyzed 

accident. As a result, no analysis 

assumptions are violated and there are no 

adverse effects on the factors that contribute 

to offsite or onsite dose as the result of an 

accident. Use of the approved methodologies 

described in the topical report being added 

to TS 5.6.5.b ensures that plant structures, 

systems, or components are maintained 

consistent with the safety analysis and 

licensing bases. Based on this evaluation, 

there is no significant increase in the 

consequences of a previously analyzed event. 

Therefore, the proposed change adding 

General Electric Nuclear Energy licensing 

topical report NEDE–32906P–A to the TS 

5.6.5.b list of documents describing approved 

methodologies for determining core operating 

limits 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 adding licensing 

topical report NEDE–32906P–A to TS 5.6.5.b, 

and the use of the analytical methods 

described therein, does not involve any 

physical alteration of plant systems, 

structures, or components, other than 

allowing for fuel and core designs in 

accordance with NRC approved 

methodologies. The proposed methodology 

continues to meet applicable criteria for core 

operating limit analysis. No new or different 

equipment is being installed. No installed 

equipment is being operated in a different 

manner. There is no alteration to the 

parameters within which the plant is 

normally operated or in the setpoints that 

initiate protective or mitigative actions. As a 

result no new failure modes are being 

introduced. 

Therefore, the proposed change adding 

General Electric Nuclear Energy licensing 

topical report NEDE–32906P–A to the TS 

5.6.5.b list of documents describing approved 

methodologies for determining core operating 

limits does not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The margin of safety is established through 

the design of the plant structures, systems, 

and components, through the parameters 

within which the plant is operated, through 

the establishment of the setpoints for the 

actuation of equipment relied upon to 

respond to an event, and through margins 

contained within the safety analyses. The 

proposed change adding General Electric 

Nuclear Energy licensing topical report 

NEDE–32906P–A to the TS 5.6.5.b list of 

documents describing approved 

methodologies for determining core operating 

limits does not impact the condition or 

performance of structures, systems, setpoints, 

and components relied upon for accident 

mitigation. The proposed change does not 

significantly impact any safety analysis 

assumptions or results. Therefore, the 

proposed change adding topical report 

NEDE–32906P–A to the TS 5.6.5.b list of 

documents describing approved 

methodologies for determining core operating 

limits does not result in 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: Steven R. Carr, 

Associate General Counsel—Legal 

Department, Progress Energy Service 

Company, LLC, Post Office Box 1551, 

Raleigh, North Carolina 27602. 

NRC Section Chief (Acting): Michael 

L. Marshall. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of amendment request:

September 22, 2004. 

Description of amendment request:

The proposed change will revise 

Columbia Generating Station’s licensing 

basis by replacing the current plantspecific reactor pressure vessel (RPV) 

material surveillance program with the 

Boiling Water Reactor Vessels and 

Internals Project (BWRVIP) Integrated 

Surveillance Program (ISP). Specifically, 

the proposed amendment would revise 

Columbia’s Final Safety Analysis Report 

(FSAR) to include participation in the 

ISP as described in the program 

document BWRVIP–86–A, ‘‘BWR Vessel 

and Internals Project Updated BWR 

Integrated Surveillance Program (ISP) 

Implementation Plan,’’ dated October 

2002. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated 

The proposed change implements an ISP 

program that meets the requirements of 10 

CFR 50, Appendix H, Paragraph III.C, 

‘‘Requirements for an Integrated Surveillance 

Program.’’ The proposed ISP program ensures 

the same level of RPV integrity as Columbia’s 

current material surveillance program. 

Implementation of the proposed ISP is not a 

precursor or initiator of any previously 

evaluated accident. No physical changes to 

Columbia Generating Station are involved 

with the proposed change. The proposed 

change will not cause the RPV or interfacing 

systems to be operated outside of any design 

limit or testing limit, and will not alter any 

assumptions or initial conditions previously 

used in evaluating the radiological 

consequences of an accident. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change revises the licensing 

basis for Columbia Generating Station to 

reflect participation in the BWRVIP ISP. The 

NRC has approved the ISP as an acceptable 

material surveillance program pursuant to 10 

CFR 50, Appendix H, paragraph III.C. No 

physical changes to the plant are associated 

with the proposed change. No changes in 

design or operation of any system, structure, 

or component will be made as a result of the 

proposed change. The ISP is an alternative 

monitoring program and cannot create a new 

failure mode or a new or different kind of 

accident from any previously evaluated. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. The proposed change does not involve 

a significant reduction in a margin of safety. 

Compliance with RPV material 

surveillance program requirements specified 

in 10 CFR 50, Appendix H and the fracture 

toughness requirements contained in 10 CFR 

62472 Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 

50, Appendix G ensure an adequate margin 

of safety exists in the fracture toughness of 

RPV beltline ferritic materials during any 

condition of normal operation, anticipated 

operational occurrence, and system 

hydrostatic tests. Implementation of the 

proposed ISP has been evaluated to meet the 

requirements of 10 CFR 50, Appendix H and 

this margin of safety is not impacted. 

Compliance with the requirements of 10 CFR 

50, Appendix G will not be affected by this 

proposed change. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Thomas C. 

Poindexter, Esq., Winston & Strawn, 

1400 L Street, NW., Washington, DC 

20005–3502. 

NRC Section Chief: Robert A. Gramm. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of amendment request:

September 27, 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 27, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated. 

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: Thomas C. 

Poindexter, Esq., Winston & Strawn, 

1400 L Street, NW., Washington, DC 

20005–3502. 

NRC Section Chief: Robert A. Gramm. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of amendment request:

September 30, 2004. 

Description of amendment request:

The proposed license amendment 

request would change the technical 

specifications and the Final Safety 

Analysis Report to revise the Columbia 

Generating Station’s licensing and 

design bases to reflect the application of 

the alternative source term (AST) 

methodology with an exception. That 

exception is TID–14844, ‘‘Calculation of 

Distance Factors for Power and Test 

Reactor Sites,’’ which will continue to 

be used as the radiation dose basis for 

equipment qualification, and radiation 

zone maps/shielding calculations. 

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 changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The alternative source term does not affect 

the design or operation of the facility in a 

manner that would impact the probability of 

an accident previously evaluated. Assumed 

performance requirements of the system 

structures and components are within 

existing design capability. The manner in 

which the systems are required to operate has 

not changed. 

Once the occurrence of an accident has 

been postulated, the new source term is an 

input to evaluate the consequences. The 

implementation of the alternative source 

term methodology has been evaluated in 

revisions to the analyses of the following 

limiting design basis accidents at Columbia 

Generating Station:

• Control Rod Drop Accident 

• Fuel Handling Accident 

• Main Steam Line Break Accident 

• Loss of Coolant Accident

This amendment request includes changes 

to the Technical Specifications based on 

assumptions in the accident analyses. The 

results of these analyses demonstrate that, 

with the requested changes, the dose 

consequences of these limiting events are 

within the regulatory limits provided by the 

NRC for use with the alternative source term. 

A new license and design basis analysis on 

secondary containment drawdown is 

provided to resolve a Justification for 

Continued Operation. The consequences, 

based on alternative source term 

methodology, remain within regulatory 

limits. This change to the licensing and 

design basis does not result in a significant 

increase in consequences. 

Alternative source term methodology has 

been applied to resolve the Unresolved 

Safety Question on control room unfiltered 

air inleakage. The accident analyses results 

show, with the increased unfiltered air 

inleakage, the control room operator doses 

remain within regulatory limits. 

Therefore, approval of the proposed 

amendment request does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The requested changes are based on 

accident analyses. System[,] structure and 

component performance assumptions 

included in the accident analyses result in 

doses within regulatory limits. Use of these 

performance assumptions does not:

• Require the installation of any new 

equipment, 

• Require the modification of any existing 

equipment, 

• Change the manner in which the 

equipment is required to be operated, 

• Assume equipment performance outside 

existing design capabilities, or 

• Require new operator actions. 

Therefore Energy Northwest application of 

the alternative source term methodology does 

not create any new accident initiators or 

precursors of a new or different kind of 

accident. 

Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62473

3. The proposed change does not involve 

a significant reduction in a margin of safety. 

The changes proposed are associated with 

the implementation of a new licensing basis 

for Columbia Generating Station. Approval of 

a basis change from the original source term 

developed in accordance with TID–14844 to 

a new alternative source term as described in 

RG [Regulatory Guide] 1.183 is requested. 

The results of the accident analyses revised 

in support of this submittal, and the 

requested Technical Specification changes, 

are subject to revised acceptance criteria. 

These analyses have been performed using 

conservative methodologies. 

Safety margins and analytical 

conservatisms have been evaluated and are 

satisfied. The analyzed accidents have been 

carefully selected and margin has been 

retained to ensure that the analyses 

adequately bound postulated event scenarios. 

The dose consequences of these limiting 

design basis accidents are within the 

acceptance criteria found in the applicable 

regulatory requirements and guidance. These 

requirements and guidance are presented in 

10 CFR 50, App. A, 10 CFR 50.67, GDC 

[General Design Criterion] 19, and RG 1.183. 

The proposed changes can be made while 

still satisfying regulatory requirements and 

review criteria, with margin. The changes 

continue to ensure that the doses at the 

exclusion area and low population zone 

boundaries, as well as the control room, are 

within the corresponding regulatory limits. 

Therefore, operation of Columbia Generating 

Station in accordance with the requested 

amendment does not involve a significant 

reduction in the margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Thomas C. 

Poindexter, Esq., Winston & Strawn, 

1400 L Street, NW., Washington, DC 

20005–3502. 

NRC Section Chief: Robert A. Gramm. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of amendment request: April 14, 

2004. 

Description of amendment request:

The proposed amendment would revise 

Technical Specifications (TSs) to allow 

a one-time interval extension of no more 

than five years for the Type A, 

Integrated Leakage Rate Test (ILRT) of 

the primary containment. The proposed 

amendment would also correct the TSs 

to remove a reference to an obsolete 

alphanumeric identifier in TS 4.7.A.2.a, 

and reformat existing text on TS Pages 

3/4.7–4 and 3/4.7–5 to improve 

consistency in its presentation. 

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. The Nuclear Regulatory 

Commission (NRC) staff has reviewed 

the licensee’s analysis against the 

standards of 10 CFR 50.92(c). The NRC 

staff’s review is presented below. 

1. Does the change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

The proposed changes do not modify 

the design or operation of the 

containment. Therefore, the proposed 

changes, therefore, will not increase the 

probability of accidents previously 

evaluated. 

The proposed extension to Type A, 

ILRT testing does not involve a 

significant increase in the consequences 

of an accident. Research documented in 

NUREG–1493 has found that Type A 

tests identify only a few potential 

containment leakage paths that also 

cannot be identified by Type B and C 

tests. The leaks that have been found by 

Type A tests have only been marginally 

above existing requirements. The 

NUREG then concluded that reducing 

the Type A testing frequency to once 

every 20 years was found to lead to an 

imperceptible increase in risk. These 

generic conclusions were confirmed by 

a plant-specific risk analysis performed 

using the current Pilgrim individual 

plant examination (IPE) internal events 

model that concluded the radiological 

consequences are low to negligible, and 

remain below regulatory limits. 

Therefore, any potential change in the 

radiological consequences is not 

considered significant. 

The proposed correction to remove 

the alphanumeric identifier (i.e., 

definition 1.U), which is no longer used 

in the TSs, from the statements 

regarding the applicability of 

surveillance frequency to leak rate tests 

is editorial in nature. Likewise, the 

proposed formatting changes to existing 

information to improve its presentation 

are also editorial in nature. Since these 

changes are administrative in nature, 

they cannot increase the probability or 

consequences of previously analyzed 

accidents. 

Therefore, since the radiological 

consequences are below the regulatory 

limits and the probability of an accident 

is unchanged, the proposed changes do 

not involve a significant increase in the 

probability or consequences of an 

accident previously evaluated. 

2. Does the proposed change create 

the possibility of a new or different kind 

of accident from any accident 

previously evaluated? 

There are no new plant operation 

modes or physical modifications being 

proposed. Therefore, the proposed 

changes do not create the possibility of 

a new or different kind of accident from 

any previously analyzed.

3. Does the change involve a 

significant reduction in a margin of 

safety? 

The proposed revisions to the TSs add 

a one-time, 5-year extension to the 

current interval of 10 years from the last 

Type A test. The NUREG–1493 generic 

study of the effects of extending 

containment leakage testing found that 

a 20-year extension in Type A leakage 

testing resulted in an imperceptible 

increase in risk to the public. The 

NUREG also found that, generically, the 

design containment leakage rate 

contributes about 0.1 percent to the 

individual risk, and that the decrease in 

Type A testing frequency would have a 

minimal affect on this risk since 95 

percent of the potential leakage paths 

are detected by Type C testing. This was 

further confirmed by a plant-specific 

risk assessment using the current 

Pilgrim IPE internal events model. 

Therefore, by meeting applicatory 

regulatory limits, any potential decrease 

in margin of safety would not be 

considered significant. 

The proposed correction to remove 

the alphanumeric identifier (i.e., 

definition 1.U), which is no longer used 

in Pilgrim TSs, from the statements 

regarding the applicability of 

surveillance frequency to leak rate tests 

is editorial in nature. Likewise, the 

proposed formatting changes to existing 

information to improve its presentation 

are also editorial in nature. As these 

changes are administrative in nature, 

the proposed changes do not involve a 

significant reduction in the margin of 

safety. 

Based on this review, it appears that 

the three standards of 10 CFR 50.92(c) 

are satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. M. Fulton, 

Esquire, Assistant General Counsel, 

Pilgrim Nuclear Power Station, 600 

Rocky Hill Road, Plymouth, 

Massachusetts, 02360–5599. 

NRC Section Chief: Daniel Collins, 

Acting. 

62474 Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2, 

Ogle County, Illinois 

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and 

2, Will County, Illinois 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

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

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois 

Date of amendment request: April 30, 

2004. 

Description of amendment request:

The proposed change allows entry into 

a mode or other specified condition in 

the applicability of a technical 

specification (TS), while in a condition 

statement and the associated required 

actions of the TS, provided the licensee 

performs a risk assessment and manages 

risk consistent with the program in 

place for complying with the 

requirements of Title 10 of the Code of 

Federal Regulations (10 CFR), Part 50, 

Section 50.65(a)(4). Limiting Condition 

for Operation (LCO) 3.0.4 exceptions in 

individual TS would be eliminated, 

several notes or specific exceptions are 

revised to reflect the related changes to 

LCO 3.0.4, and Surveillance 

Requirement (SR) 3.0.4 is revised to 

reflect the LCO 3.0.4 allowance. 

This change was proposed by the 

industry’s Technical Specification Task 

Force (TSTF) and is designated TSTF–

359. The Nuclear Regulatory 

Commission (NRC) staff issued a notice 

of opportunity for comment in the 

Federal Register on August 2, 2002 (67 

FR 50475), on possible amendments 

concerning TSTF–359, including a 

model safety evaluation and model no 

significant hazards consideration 

(NSHC) determination, using the 

consolidated line item improvement 

process. The NRC staff subsequently 

issued a notice of availability of the 

models for referencing in license 

amendment applications in the Federal 

Register on April 4, 2003 (68 FR 16579). 

The licensee affirmed the applicability 

of the following NSHC determination in 

its application dated April 30, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of NSHC is 

presented below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change allows entry into a 

mode or other specified condition in the 

applicability of a TS, while in a TS condition 

statement and the associated required actions 

of the TS. Being in a TS condition and the 

associated required actions is not an initiator 

of any accident previously evaluated. 

Therefore, the probability of an accident 

previously evaluated is not significantly 

increased. The consequences of an accident 

while relying on required actions as allowed 

by proposed LCO 3.0.4, are no different than 

the consequences of an accident while 

entering and relying on the required actions 

while starting in a condition of applicability 

of the TS. Therefore, the consequences of an 

accident previously evaluated are not 

significantly affected by this change. The 

addition of a requirement to assess and 

manage the risk introduced by this change 

will further minimize possible concerns. 

Therefore, this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Previously 

Evaluated 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed). 

Entering into a mode or other specified 

condition in the applicability of a TS, while 

in a TS condition statement and the 

associated required actions of the TS, will 

not introduce new failure modes or effects 

and will not, in the absence of other 

unrelated failures, lead to an accident whose 

consequences exceed the consequences of 

accidents previously evaluated. The addition 

of a requirement to assess and manage the 

risk introduced by this change will further 

minimize possible concerns. Thus, this 

change does not create the possibility of a 

new or different kind of accident from an 

accident previously evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in a Margin 

of Safety

The proposed change allows entry into a 

mode or other specified condition in the 

applicability of a TS, while in a TS condition 

statement and the associated required actions 

of the TS. The TS allow operation of the 

plant without the full complement of 

equipment through the conditions for not 

meeting the TS LCO. The risk associated with 

this allowance is managed by the imposition 

of required actions that must be performed 

within the prescribed completion times. The 

net effect of being in a TS condition on the 

margin of safety is not considered significant. 

The proposed change does not alter the 

required actions or completion times of the 

TS. The proposed change allows TS 

conditions to be entered, and the associated 

required actions and completion times to be 

used in new circumstances. This use is 

predicated upon the licensee’s performance 

of a risk assessment and the management of 

plant risk. The change also eliminates current 

allowances for utilizing required actions and 

completion times in similar circumstances, 

without assessing and managing risk. The net 

change to the margin of safety is 

insignificant. Therefore, this change does not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Thomas S. 

O’Neill, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Section Chief: Gene Suh. 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station, Units 1 

and 2, Montgomery County, 

Pennsylvania 

Date of amendment request: May 20, 

2004. 

Description of amendment request:

The proposed changes would modify 

the Limerick Generating Station (LGS), 

Units 1 and 2, Technical Specifications 

(TSs) to support activation of the trip 

outputs of the previously-installed 

Oscillation Power Range Monitor 

(OPRM) portion of the Power Range 

Neutron Monitoring (PRNM) system. 

Specifically, the proposed changes 

would revise LGS TS 2.2.1, ‘‘Reactor 

Protection System Instrumentation 

Setpoints,’’ TS 3/4.3.1, ‘‘Reactor 

Protection System Instrumentation,’’ TS 

3/4.3.6, ‘‘Control Rod Block 

Instrumentation,’’ TS 3/4.4.1, 

‘‘Recirculation System’’ and their 

associated Bases. 

The proposed changes would also 

revise TS 6.9.1, ‘‘Routine Reports,’’ and 

delete interim corrective action 

requirements from the Recirculation 

System TS. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. This modification has no 

impact on any of the previously installed 

PRNM functions. Plant operation in portions 

of the former restricted zone may potentially 

cause a marginal increase in the probability 

of occurrence of an instability event. This 

potential increase in probability is acceptable 

Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62475

because the OPRM Upscale Function will 

automatically detect the condition and 

initiate a reactor scram before the Minimum 

Critical Power Ratio (MCPR) Safety Limit is 

reached. Consequences of the potential 

instability event are reduced because of the 

more reliable automatic detection and 

suppression of an instability event, and 

elimination of dependence on the manual 

operator actions. 

The change to align the operability 

requirements for the Intermediate Range 

Monitor (IRM) rod block function with those 

for the corresponding IRM Reactor Protection 

System (RPS) functions affects only the rod 

block function. The justification for the 

change to IRM RPS function (done with the 

original PRNM modification) concluded that 

the RPS change would not increase the 

probability of occurrence of an accident 

previously evaluated; therefore, changing the 

associated rod block to align with those 

requirements would not do so either. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability of consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. The modification replaces 

procedural actions that were established to 

avoid operating conditions where reactor 

instabilities might occur with an NRC 

approved automatic detect and suppress 

function. 

Potential failures in the OPRM Upscale 

Function could result in either failure to take 

the required mitigating action or an 

unintended reactor scram. These are the 

same potential effects of failure of the 

operator to take the appropriate action under 

the current procedural directions. The net 

effect of the modification changes the method 

by which an instability event is detected and 

by which mitigating action is initiated, but 

does not change the type of stability event 

that could occur. The effects of failure of the 

OPRM equipment are limited to reduced or 

failed mitigation, but such failure cannot 

cause an instability event or other type of 

accident. 

The change to align the operability 

requirements for the IRM rod block function 

with those for the corresponding IRM RPS 

functions affects only the rod block function. 

The justification for the change to IRM RPS 

function (done with the original PRNM 

modification) concluded that the RPS change 

could not create the possibility of a new type 

of accident; therefore, changing the 

associated rod block to align with those 

requirements would not do so either. 

Therefore, since no radiological barrier will 

be challenged as a result of activating the 

OPRM Upscale Function, it is concluded that 

the proposed changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. The current safety analysis 

assumes that the existing procedural actions 

are adequate to prevent an instability event. 

As a result, there is currently no quantitative 

or qualitative assessment of an instability 

event with respect to its impact on the MCPR 

Safety Limit. 

The OPRM Upscale function is being 

implemented to automate the detection (via 

direct measurement of neutron flux) and 

subsequent suppression (via scram) of an 

instability event prior to exceeding the MCPR 

Safety Limit. The OPRM Upscale function 

provides a trip output of the same type as 

currently used for the Average Power Range 

Monitor (APRM). Its failure modes and types 

are identical to those for the present APRM 

output. Currently, the MCPR Safety Limit is 

not challenged by an instability event since 

the event is ‘‘mitigated’’ by manual means via 

the procedural actions, which prevent plant 

operating conditions where an instability 

event is possible. In both methods of 

mitigation (manual and automated), the 

margin of safety associated with the MCPR 

Safety Limit is still maintained. 

Therefore, based on the fact that the MCPR 

Safety Limit will still be enforced, 

implementation of the OPRM Upscale 

function in place of the existing manual 

actions does not reduce the margin of safety. 

The IRM rod block function is not 

considered in any safety analysis. As a result, 

its failure will not affect the margin of safety. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Thomas S. 

O’Neill, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Section Chief: Daniel S. Collins, 

Acting. 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station, Units 1 

and 2, Montgomery County, 

Pennsylvania 

Date of amendment request: June 1, 

2004. 

Description of amendment request:

The proposed changes would relocate 

the operability and surveillance 

requirements for the reactor coolant 

system safety/relief valve position 

instrumentation from the Limerick 

Generating Station (LGS) Technical 

Specifications (TSs) to the LGS 

Technical Requirements Manual (TRM) 

and plant procedures. Specifically, the 

changes would relocate TSs 3.4.2.c, 

4.4.2.1, and the associated footnotes to 

the TRM. Additionally, the ‘‘Safety/

Relief Valve Position Indicators’’

instrumentation would be relocated 

from Tables 3.3.7.5–1 and 4.3.7.5–1 of 

TSs 3.3.7.5 and 4.3.7.5, respectively to 

the TRM. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. The failure of the safety/

relief valve (SRV) position instrumentation is 

not assumed to be an initiator of any 

analyzed event in the [Updated Final Safety 

Analysis Report] UFSAR. The proposed 

changes do not alter the physical design of 

the SRVs or any other plant structure, 

system, or component. The changes would 

remove the [SRV] position indicator 

operability and surveillance requirements 

from the LGS [TSs], and incorporate 

requirements verbatim for this 

instrumentation into a licensee-controlled 

document under the control of 10 CFR 50.59. 

The proposed changes conform to NRC 

regulatory guidance regarding the content of 

plant [TSs] as identified in regulation 10 CFR 

50.36, and NRC publication NUREG–1433, 

‘‘Standard Technical Specifications—General 

Electric Plants, BWR/4.’’

Therefore, this proposed amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. The proposed changes do 

not alter the physical design, safety limits, or 

safety analysis assumptions, associated with 

the operation of the plant. Accordingly, the 

proposed changes do not introduce any new 

accident initiators, nor do they reduce or 

adversely affect the capabilities of any plant 

structure or system in the performance of 

their safety function. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. This instrumentation is not 

needed for manual operator actions necessary 

for safety systems to accomplish their safety 

function for the design basis accident events. 

The instrumentation provides only alarm and 

SRV position indication, and does not 

provide an input to any automatic trip 

function. Several diverse means are available 

to monitor SRV position, and operability and 

surveillance requirements will be established 

in a licensee-controlled document to assure 

the reliability of SRV position monitoring 

capability. Changes to these requirements 

will be subject to the controls of regulation 

10 CFR 50.59, providing the appropriate level 

of regulatory control. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety.

62476 Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Thomas S. 

O’Neill, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Section Chief: Daniel S. Collins, 

Acting. 

Exelon Generation Company, LLC, and 

PSEG Nuclear LLC, Docket Nos. 50–277 

and 50–278, Peach Bottom Atomic 

Power Station, Units 2 and 3, York and 

Lancaster Counties, Pennsylvania 

Date of application for amendments:

June 24, 2004. 

Description of amendment request:

The proposed amendment would revise 

Technical Specification 3.1.8, ‘‘Scram 

Discharge Volume (SDV) Vent and Drain 

Valves,’’ to allow a vent or drain line 

with one inoperable valve to be isolated 

instead of requiring the valve to be 

restored to Operable status within 7 

days. Other changes included in the 

application are addressed in a separate 

notice. 

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on February 24, 2003 (68 FR 

8637), on possible amendments to revise 

the action for one or more SDV vent or 

drain lines with an inoperable valve, 

including a model safety evaluation and 

model no significant hazards 

consideration (NSHC) determination, 

using the consolidated line-item 

improvement process. The NRC staff 

subsequently issued a notice of 

availability of the models for referencing 

in license amendment applications in 

the Federal Register on April 15, 2003 

(68 FR 18294). The licensee affirmed the 

applicability of the model NSHC 

determination in its application dated 

June 24, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated

A change is proposed to allow the affected 

SDV vent and drain line to be isolated when 

there are one or more SDV vent or drain lines 

with one valve inoperable instead [of] 

requiring the valve to be restored to operable 

status within 7 days. With one SDV vent or 

drain valve inoperable in one or more lines, 

the isolation function would be maintained 

since the redundant valve in the affected line 

would perform its safety function of isolating 

the SDV. Following the completion of the 

required action, the isolation function is 

fulfilled since the associated line is isolated. 

The ability to vent and drain the SDVs is 

maintained and controlled through 

administrative controls. This requirement 

assures the reactor protection system is not 

adversely affected by the inoperable valves. 

With the safety functions of the valves being 

maintained, the probability or consequences 

of an accident previously evaluated are not 

significantly increased. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Accident 

Previously Evaluated 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. Thus, this change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The proposed change ensures that the 

safety functions of the SDV vent and drain 

valves are fulfilled. The isolation function is 

maintained by redundant valves and by the 

required action to isolate the affected line. 

The ability to vent and drain the SDVs is 

maintained through administrative controls. 

In addition, the reactor protection system 

will prevent filling of an SDV to the point 

that it has insufficient volume to accept a full 

scram. Maintaining the safety functions 

related to isolation of the SDV and insertion 

of control rods ensures that the proposed 

change does not involve a significant 

reduction in the margin of safety.

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for Licensee: Thomas S. 

O’Neill, Associate and General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Section Chief: Daniel Collins, 

Acting. 

Indiana Michigan Power Company, 

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

C. Cook Nuclear Plant, Units 1 and 2, 

Berrien County, Michigan 

Date of amendment requests:

September 21, 2004. 

Description of amendment requests:

The proposed amendments would 

extend the allowed outage times from 72 

hours to 14 days for an inoperable 

emergency diesel generator, an 

inoperable component cooling water 

system loop, an inoperable essential 

service water system loop, or an 

inoperable alternate offsite power 

circuit (69 kilovolt circuit). The 

proposed amendments would also 

change formats of the affected technical 

specification pages to improve their 

appearance but not alter any 

requirements. 

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 of 

occurrence or consequences of an accident 

previously evaluated? 

Response: No. 

The proposed changes to the Technical 

Specifications (TS) will extend the allowed 

outage time (AOT) for a single inoperable 

emergency diesel generator (EDG), one 

inoperable component cooling water (CCW) 

or essential service water (ESW) loop, or an 

inoperable 69 kilovolt (kV) offsite circuit 

from the current limit of 72 hours to 14 days. 

An independent alternating current (AC) 

power source consisting of two supplemental 

diesel generators (SDGs) will be installed to 

support the extended AOTs for the EDGs and 

the CCW and ESW systems. The SDGs will 

supply power to required safe shutdown 

loads in the affected unit. 

The EDGs are backup AC power sources 

designed to power safe shutdown systems in 

the event of a loss of offsite power. As such, 

the EDGs are not initiators for any accident 

previously evaluated. The CCW and ESW 

systems provide cooling water to safetyrelated components. This is a support 

function, and malfunctions of the CCW and 

ESW systems are not initiators of any 

accidents previously analyzed. The 69 kV 

circuit is an alternate offsite power supply 

that must be manually connected by the 

control room operators to provide power to 

safety-related buses upon loss of the 

preferred 34.5 kV offsite power source. As 

such, the 69 kV circuit is not an initiator for 

any accident previously evaluated. The AOT 

extension for an inoperable EDG, a CCW or 

ESW loop, or 69 kV circuit does not 

introduce any failure mechanisms that would 

initiate a previously analyzed accident. 

Therefore, the proposed change permitting 

extension of the AOTs for the EDG, ESW, 

CCW, and 69 kV systems do not result in a 

significant increase in the probability of a 

previously evaluated accident.

The potential effect of the proposed change 

on the consequences of a previously 

evaluated accident has been considered. 

There are two EDGs per unit, and only one 

EDG per unit is required to fulfill the onsite 

AC power system safety function. During the 

extended AOT, the redundant EDG will be 

available to provide AC power to safetyrelated components. There are two CCW 

loops per unit, and only one CCW loop per 

unit is required to fulfill the CCW system 

safety function. During the extended AOT, 

the redundant CCW loop will be available to 

provide cooling water to safety-related 

components. There are two ESW loops per 

unit, and only one ESW loop per unit is 

Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62477

required to fulfill the ESW system safety 

function for the affected unit. During the 

extended AOT, the redundant ESW loop will 

be available to provide cooling water to the 

safety-related components. The 69 kV offsite 

circuit is the alternate offsite power source. 

Only one offsite power source is required to 

fulfill the offsite power system safety 

function. During an extended AOT, the 

preferred offsite source will be available. 

Thus, the systems affected by the proposed 

amendment will still be capable of 

performing the safety functions needed to 

mitigate the consequences of an accident as 

previously evaluated. 

The format changes improve appearance, 

but do not affect any requirements. 

Therefore, the proposed change will not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change consists of increasing 

the AOTs allowed by TS for the EDG, CCW, 

ESW, and 69 kV systems. Extending existing 

AOTs, does not result in operation of the 

plant in any new or different manner, nor 

does it create any new accident precursors. 

The format changes improve appearance, but 

do not affect any requirements. 

Therefore, the proposed change will not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The margins of safety are established 

through design parameters, operating 

parameters, and the setpoints at which 

automatic actions are initiated. The proposed 

change does not adversely affect any design 

or operating parameter or any setpoint used 

in the deterministic accident analyses to 

establish the margin of safety. Probabilistic 

risk assessment methods were used to 

evaluate the risked-based margins of safety 

for the proposed change. The results of these 

evaluations indicated the proposed AOT 

extensions combined with installation of 

additional on-site electrical power supplies 

results in a net risk reduction. The format 

changes improve appearance, but do not 

affect any requirements. 

Therefore, the proposed change will not 

create a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: David W. 

Jenkins, Esq., 500 Circle Drive, 

Buchanan, MI 49107. 

NRC Section Chief: L. Raghavan. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request:

September 7, 2004. 

Description of amendment request:

The proposed amendment would delete 

Technical Specification (TS) 5.9.1b, 

‘‘Annual Occupational Exposure 

Report,’’ and TS 5.9.1c, ‘‘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 7, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

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: James R. 

Curtiss, Esq., Winston & Strawn, 1400 L 

Street, NW., Washington, DC 20005–

3502. 

NRC Section Chief: Robert A. Gramm. 

STP Nuclear Operating Company, 

Docket No. 50–498, South Texas Project, 

Unit 1, Matagorda County, Texas

Date of amendment request:

September 30, 2004. 

Description of amendment request:

The amendment would change 

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. 

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 block valve for the pressurizer power 

operated relief valve is not a potential 

accident initiator. Therefore, not requiring a 

surveillance of the block valve while it is 

being used to isolate its associated power 

operated relief valve will not increase the 

probability of an accident previously 

evaluated. Not requiring the surveillance of 

the block valve may slightly reduce the 

probability of a loss of coolant accident from 

a stuck open power operated relief valve 

since it will eliminate the challenge to the 

power operated relief valve from the pressure 

transient that results from cycling the block 

valve. 

If pressurizer spray is not available or is 

not effective, either one of the two 

pressurizer power operated relief valves may 

be manually actuated to depressurize the 

reactor coolant system to mitigate the 

consequences of a steam generator tube 

rupture. Not performing the surveillance on 

the block valve is not relevant to the primary 

system for depressurizing the reactor coolant 

system (pressurizer spray). The block valves 

have been demonstrated by operating 

experience to be reliable and are also subject 

to the motor-operated valve testing program. 

Consequently, the proposed change does not 

significantly reduce the confidence that the 

block valve can be opened to permit manual 

actuation of the power operated relief valve 

to depressurize the reactor coolant system to 

mitigate an accident. Therefore, the proposed 

change does not involve a significant 

62478 Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 

increase in the consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different accident 

from any accident previously evaluated? 

Response: No. 

The proposed change only affects the 

performance of the surveillance test for the 

block valve and does not introduce any 

operating configurations not previously 

evaluated. 

Therefore, STPNOC concludes the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change to the surveillance 

requirement for the block valve for the 

pressurizer power operated relief valve does 

not affect the assumptions in any accident 

analyses. There are no changes in plant 

performance parameters associated with the 

proposed change to the surveillance 

requirement for the block valve. Therefore, 

STPNOC concludes 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 standards of 

10 CFR 50.92(c) are satisfied. Therefore, 

the NRC staff proposes to determine that 

the request for amendment involves no 

significant hazards consideration. 

Attorney for licensee: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Section Chief: Michael K. Webb, 

Acting. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request:

September 27, 2004. 

Description of amendment request:

The proposed amendment would 

change the Technical Specifications 

(TS) to provide consistency between 

Surveillance Requirement (SR) 4.7.1.6 

and TS 3.3.5.1 regarding atmospheric 

steam relief valve instrumentation 

controls. The proposed amendment 

would also correct editorial errors in TS 

3.7.1.6. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

significant increase in the probability or 

consequences of a previously evaluated 

accident. The first proposed change only 

clarifies when SR 4.7.1.6 for the automatic 

controls of the atmospheric steam relief valve 

is applicable. The applicability is already 

established in TS 3.3.5.1 and meets the safety 

analysis. The second proposed change is 

editorial. 

2. Does the proposed change create the 

possibility of a new or different accident 

from any accident previously evaluated? 

Response: No. 

The first proposed change does not create 

the possibility of a new or different accident 

from any previously evaluated. The proposed 

change only clarifies when SR 4.7.1.6 for the 

automatic controls of the atmospheric steam 

relief valve is applicable. The applicability is 

already established in TS 3.3.5.1 and meets 

the safety analysis. The second proposed 

change is editorial. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The first proposed change does not involve 

a significant reduction in the margin of 

safety. The proposed change only clarifies 

when SR 4.7.1.6 for the automatic controls of 

the atmospheric steam relief valve is 

applicable. The applicability is already 

established in TS 3.3.5.1 and meets the safety 

analysis. The second proposed change is 

editorial.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the standards of 

10 CFR 50.92(c) are satisfied. Therefore, 

the NRC staff proposes to determine that 

the request for amendments involves no 

significant hazards consideration.

Attorney for licensee: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Section Chief: Michael K. Webb, 

Acting. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request:

September 30, 2004. 

Description of amendment request:

The proposed amendment would delete 

Technical Specification (TS) 6.9.1.2, 

‘‘Occupational Radiation Exposure 

Report,’’ and TS 6.9.1.5, ‘‘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 30, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

Criterion 1—Does the Proposed Change 

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—Does the Proposed Change 

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—Does the Proposed Change 

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: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Section Chief: Michael K. Webb, 

Acting. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request:

September 9, 2004. 

Brief description of amendments: The 

proposed change will revise the 

surveillance requirement (SR) 3.6.6.8 

Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62479

frequency of every 10 years. Instead, the 

proposed change to SR 3.6.6.8 will 

require verification that spray nozzles 

are unobstructed following maintenance 

that could result in nozzle blockage. 

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. The NRC staff has 

reviewed the Licensee’s analysis against 

the standards of 10 CFR 50.92(c). The 

NRC staff’s review is presented below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The Containment Spray System is not 

considered an initiator of any analyzed 

event. The proposed change does not 

have a detrimental impact on the 

integrity of any plant structure, system, 

or component that may initiate an 

analyzed event. The proposed change 

will not alter the operation or otherwise 

increase the failure probability of any 

plant equipment that can initiate an 

analyzed accident. This change does not 

affect the plant design. There is no 

increase in the likelihood of formation 

of significant corrosion products. Due to 

their location at the top of the 

containment, introduction of foreign 

material into the spray headers is 

unlikely. Foreign material introduced 

during maintenance activities would be 

the most likely source for obstruction, 

and verification following such 

maintenance would confirm the nozzles 

remain unobstructed. 

Consequently, there is no significant 

increase in the probability of an 

accident previously evaluated. 

The Containment Spray System is 

designed to address the consequences of 

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

Containment Spray System is capable of 

performing its function effectively with 

the single failure of any active 

component in the system, any of its 

subsystems, or any of its support 

systems. A plugged nozzle would have 

negligible impact on the capability of 

the Containment Spray System to 

respond to a Loss of Coolant Accident. 

Therefore, the consequences of an 

accident previously evaluated are not 

significantly affected by the proposed 

change. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated?

Response: No. 

The proposed change will not 

physically alter the plant (no new or 

different type of equipment will be 

installed) or change the methods 

governing normal plant operation. 

Therefore, the proposed change does 

not create the possibility of a new or 

different kind of accident from any 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of 

safety? 

Response: No. 

The system is not susceptible to 

corrosion-induced obstruction or 

obstruction from sources external to the 

system. Maintenance activities that 

could introduce foreign material into 

the system would require subsequent 

verification to ensure there is no nozzle 

blockage. The spray header nozzles are 

expected to remain unblocked and 

available in the event that the safety 

function is required. Therefore, the 

capacity of the system would remain 

unaffected. 

Therefore the proposed change does 

not involve a reduction in a margin of 

safety. 

Based on this review, it appears that 

the three standards of 10 CFR 50.92(c) 

are satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Section Chief: Mohan Thadani, 

Acting Chief. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request:

September 10, 2004. 

Brief description of amendments: 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 10, 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:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

The proposed change eliminates the 

Technical Specifications (TSs) reporting 

requirements to provide a monthly operating 

report of shutdown experience and operating 

statistics if the equivalent data is submitted 

using an industry electronic database. It also 

eliminates the TS reporting requirement for 

an annual occupational radiation exposure 

report, which provides information beyond 

that specified in NRC regulations. The 

proposed change involves no changes to 

plant systems or accident analyses. As such, 

the change is administrative in nature and 

does not affect initiators of analyzed events 

or assumed mitigation of accidents or 

transients. Therefore, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

The proposed change does not involve a 

physical alteration of the plant, add any new 

equipment, or require any existing 

equipment to be operated in a manner 

different from the present design. Therefore, 

the proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

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: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Section Chief: Michael Webb, 

Acting. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

62480 Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 

page cited. This notice does not extend 

the notice period of the original notice. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of amendment request: April 1, 

2004.

Brief description of amendment 

request: The proposed amendment 

would allow entry into a mode or other 

specified condition in the applicability 

of a technical specification (TS), while 

in a condition statement and the 

associated required actions of the TS, 

provided the licensee performs a risk 

assessment and manages risk consistent 

with the program in place for complying 

with the requirements of Title 10 of the 

Code of Federal Regulations, Part 50, 

Section 50.65(a)(4). Limiting Condition 

for Operation (LCO) 3.0.4 exceptions in 

individual TSs would be eliminated, 

and Surveillance Requirement 3.0.4 

revised to reflect the LCO 3.0.4 

allowance. 

Date of publication of individual 

notice in Federal Register: August 24, 

2004 (69 FR 52037). 

Expiration date of individual notice:

October 23, 2004. 

Florida Power and Light Company, et 

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

Lucie Plant, Unit No. 1, and Unit No. 2, 

St. Lucie County, Florida 

Date of amendment request:

November 21, 2003. 

Description of amendment request:

Revise Technical Specifications to 

eliminate certain pressure sensor 

response time testing requirements as 

discussed in the Combustion 

Engineering Owners Group Topical 

Report NPSD–1167, Revision 2, 

Elimination of Pressure Sensor 

Response Time Testing Requirements.’’

Date of publication of individual 

notice in the Federal Register:

September 28, 2004 (69 FR 57975). 

Expiration date of individual notice:

November 29, 2004. 

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 TS 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 publication of individual 

notice in Federal Register: October 6, 

2004 (69 FR 59969). 

Expiration date of individual notice:

October 20, 2004 (Comment); December 

6, 2004 (Hearing). 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

Systems (ADAMS) Public Electronic 

Reading Room on the internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the NRC 

Public Document Room (PDR) Reference 

staff at 1–800–397–4209, (301) 415–

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

AmerGen Energy Company, LLC, et al., 

Docket No. 50–219, Oyster Creek 

Nuclear Generating Station, Ocean 

County, New Jersey 

Date of application for amendment:

February 27, 2004, as supplemented by 

letter dated August 11, 2004. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications, relocating the average 

power range monitor flux scram setting 

and rod block setting from the to the 

Core Operating Limits Report. 

Date of Issuance: October 4, 2004.

Effective date: October 4, 2004 and 

shall be implemented within 60 days of 

issuance. 

Amendment No.: 248. 

Facility Operating License No. DPR–

16: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: April 13, 2004 (69 FR 19563). 

The supplement dated August 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 Nuclear Regulatory Commission 

(NRC) staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of this amendment is contained in a 

Safety Evaluation dated October 4, 2004. 

No significant hazards consideration 

comments received: No. 

AmerGen Energy Company, LLC, et al., 

Docket No. 50–219, Oyster Creek 

Nuclear Generating Station, Ocean 

County, New Jersey 

Date of application for amendment:

March 19, 2004. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications, changing the 

surveillance requirements associated 

with control rod scram time testing. 

Specifically, the amendment modified 

the conditions under which scram time 

testing of control rods is required, and 

added a requirement to perform such 

testing on a defined portion of control 

rods at a specified frequency during the 

operating cycle. 

Date of Issuance: October 4, 2004. 

Effective date: October 4, 2004 and 

shall be implemented within 60 days of 

issuance. 

Amendment No.: 249. 

Facility Operating License No. DPR–

16: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: April 27, 2004 (69 FR 22878). 

The Commission’s related evaluation 

of this amendment is contained in a 

Safety Evaluation dated October 4, 2004. 

Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62481

No significant hazards consideration 

comments received: No. 

Carolina Power & Light Company, et al., 

Docket No. 50–400, Shearon Harris 

Nuclear Power Plant, Unit 1, Wake and 

Chatham Counties, North Carolina 

Date of application for amendment:

May 5, 2004, as supplemented August 6, 

2004. 

Brief description of amendment: This 

amendment adds a reference to the 

American Society of Mechanical 

Engineers Code for Operation and 

Maintenance of Nuclear Power Plants in 

Technical Specification Surveillance 

Requirement 4.0.5.a for the snubbers. 

Date of issuance: October 1, 2004. 

Effective date: October 1, 2004. 

Amendment No. 117. 

Facility Operating License No. NPF–

63. Amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: June 22, 2004 (69 FR 34697). 

The August 6, 2004, supplement 

contained clarifying information only 

and did not change the initial proposed 

no significant hazards consideration 

determination or expand the scope of 

the initial application. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 1, 2004. 

No significant hazards consideration 

comments received: No. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of application for amendment:

February 13, 2003, as supplemented 

July 8, 2003, December 12, 2003, June 4, 

2004, July 30, 2004, and September 16, 

2004.

Brief description of amendment: The 

amendment approves the use of an 

alternative source term methodology in 

accordance with Title 10 of the Code of 

Federal Regulations, Part 50, Section 

50.67, based on a reevaluation of the 

design-basis loss-of-coolant and fuel 

handling accidents. In addition to 

related design-basis changes, the 

amendment revises the Technical 

Specifications to (1) permit an increase 

in the allowable leak rate for the main 

steam isolation valves (MSIVs), (2) 

increase the allowable secondary 

containment bypass leakage, (3) delete 

the MSIV leakage control system, and 

(4) increase the allowed secondary 

containment draw-down time. 

Date of issuance: September 28, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment No.: 160. 

Facility Operating License No. NPF–

43: Amendment revises the Technical 

Specifications and authorizes changes to 

the Updated Final Safety Analysis 

Report. 

Date of initial notice in Federal 

Register: May 27, 2003 (68 FR 28847). 

The July 8, 2003, December 12, 2003, 

June 4, 2004, July 30, 2004, and 

September 16, 2004, supplemental 

letters provided additional clarifying 

information that was within the scope of 

the original application and did not 

change the Nuclear Regulatory 

Commission staff’s initial proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 28, 

2004. 

No significant hazards consideration 

comments received: No. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of application for amendment:

January 30, 2004. 

Brief description of amendment: The 

amendment revises technical 

specification 3.3.6.2, ‘‘Secondary 

Containment Isolation Instrumentation, 

Condition C, to add the words ‘‘not 

met’’ to the end of the phrase, ‘‘Required 

Action and associated Completion 

Time.’’ The omission of the words ‘‘not 

met’’ was an oversight during the 

change to Improved Standard Technical 

Specifications, NUREG 1433. 

Date of issuance: January 30, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 161. 

Facility Operating License No. NPF–

43: Amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: June 22, 2004, (69 FR 34698). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 7, 2004. 

No significant hazards consideration 

comments received: No. 

Entergy Gulf States, Inc., and Entergy 

Operations, Inc., Docket No. 50–458, 

River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: March 4, 

2004, as supplemented by letter dated 

June 16, 2004. 

Brief description of amendment: The 

amendment revises the Technical 

Specification requirements by 

eliminating the requirements associated 

with hydrogen recombiners and 

hydrogen monitors. These changes 

support implementation of the revisions 

to Title 10 of the Code of Federal 

Regulations (10 CFR), Section 50.44, 

‘‘Standards for Combustible Gas Control 

System In Light-Water-Cooled Power 

Reactors.’’ A notice of availability of this 

TS improvement was published in the 

Federal Register on September 25, 2003 

(68 FR 55416). 

Date of issuance: October 4, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days of Issuance. 

Amendment No.: 142. 

Facility Operating License No. NPF–

47: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: May 11, 2004 (69 FR 26187). 

The supplement dated June 16, 2004, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 4, 2004. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant, Oswego County, 

New York 

Date of application for amendment:

July 28, 2003, as supplemented on May 

20, 2004. 

Brief description of amendment: The 

amendment revised Technical 

Specification 5.5.6, ‘‘Primary 

Containment Leakage Rate Testing 

Program,’’ to allow a one-time extension 

of the interval between the Type A, 

integrated leakage rate tests, from 10 

years to no more than 15 years. 

Therefore, the first Type A test 

performed after the March 7, 1995, test 

shall be performed no later than March 

7, 2010. 

Date of issuance: September 28, 2004. 

Effective date: As of the date of 

issuance to be implemented within 30 

days. 

Amendment No.: 279. 

Facility Operating License No. DPR–

59: Amendment revised the Technical 

Specifications.

Date of initial notice in Federal 

Register: July 27, 2004 (69 FR 44696). 

The supplement dated May 20, 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 amendment is contained in a 

62482 Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 

Safety Evaluation dated September 28, 

2004. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2, 

Ogle County, Illinois Docket Nos. STN 

50–456 and STN 50–457, Braidwood 

Station, Unit Nos. 1 and 2, Will County, 

Illinois. 

Date of application for amendments:

May 21, 2004. 

Brief description of amendments: The 

amendments revise the technical 

specifications to add an additional 

reference as an acceptable method for 

determining the reactor pressure vessel 

pressure-temperature limits. 

Date of issuance: October 4, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 139/139, 132/132. 

Facility Operating License Nos. NPF–

37, NPF–66, NPF–72 and NPF–77: The 

amendments revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 3, 2004. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 4, 2004. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

Date of application for amendments:

January 15, 2004, and supplemented on 

June 22, 2004. 

Brief description of amendments: The 

amendments allow for a one-time 

deferral of the Dresden, Units 2 and 3, 

Appendix J, Type A, Integrated Leakage 

Rate Test (ILRT) to no later than 

February 27, 2011, and July 13, 2009, 

respectively. 

Date of issuance: October 13, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 210/202. 

Facility Operating License Nos. DPR–

19 and DPR–25: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: March 16, 2004. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 13, 

2004. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

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

County Station, Units 1 and 2, LaSalle 

County, Illinois; Docket Nos. 50–237 

and 50–249, Dresden Nuclear Power 

Station, Units 2 and 3, Grundy County, 

Illinois; Docket Nos. 50–254 and 50–

265, Quad Cities Nuclear Power Station, 

Units 1 and 2, Rock Island County, 

Illinois 

Date of application for amendments:

November 3, 2003. 

Brief description of amendments: The 

amendments revised Technical 

Specification (TS) 3.4.1, ‘‘Recirculation 

Loops Operating,’’ by adding a limiting 

condition for operation requirement that 

the linear heat generation rate (LHGR) 

limits shall be modified for single 

recirculation loop operation as specified 

in the Core Operating Limits Report. 

The associated TS Bases are also revised 

to reflect the new LHGR limit 

requirement. 

Date of issuance: October 4, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 167, 153, 209, 201, 

221, 216. 

Facility Operating License Nos. NPF–

11, NPF–18, DPR–19, DPR–25, DPR–29 

and DPR–30: The amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: January 6, 2004 (69 FR 694). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 4, 2004. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–334, 

Beaver Valley Power Station, Unit No. 1, 

Beaver County, Pennsylvania 

Date of application for amendment:

January 27, 2004, as supplemented May 

27, 2004. 

Brief description of amendment: The 

amendment revised TS 3.4.5 to allow a 

one-cycle use of Westinghouse leaklimiting Alloy 800 SG tube sleeves as an 

acceptable SG tube repair. Specifically, 

surveillance requirements 4.4.5.4.a.6 

and 4.4.5.4.a.9 are revised to list the 

Westinghouse leak-limiting Alloy 800 

sleeves as an acceptable SG tube 

sleeving method in addition to the 

currently approved Westinghouse laser 

welded sleeves and the former ABB 

Combustion Engineering tungsten inert 

gas welded sleeves. 

Date of issuance: October 5, 2004. 

Effective date: Within 60 days of the 

date of issuance and shall include the 

licensee commitments contained in the 

licensee letters of January 27 and May 

27, 2004. The commitments shall 

remain in effect for the authorized 

period of sleeving with Westinghouse 

Alloy 800 tubes, i.e., Cycle 17.

Amendment No: 260. 

Facility Operating License No. DPR–

66: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: March 16, 2004 (69 FR 

12369). 

The supplement dated May 27, 2004, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the Nuclear 

Regulatory Commission (NRC) staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register on 

March 16, 2004 (69 FR 12369). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 5, 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:

October 17, 2003. 

Brief description of amendments:

These amendments revised the action 

requirements in TS 3/4 6.3 to more 

clearly define the action requirements 

for inoperable containment isolation 

valves (CIVs). The amendments also 

allowed under administrative control, 

the intermittent unisolating of 

penetration flow paths which have 

previously been isolated per the action 

requirements. The amendments also 

allowed the use of check valves as an 

isolation device, and an increase in the 

allowed outage time to 72 hours for 

CIVs associated with closed systems 

inside containment. The amendments 

also deleted existing surveillance 

requirements (SRs) and provided new 

SRs similar to those in the Improved 

Standard Technical Specifications. 

Date of issuance: October 5, 2004. 

Effective date: As of date of issuance 

and shall be implemented within 60 

days. 

Amendment Nos.: 261 and 143. 

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: November 25, 2003 (68 FR 

66136). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 5, 2004. 

No significant hazards consideration 

comments received: No. 

Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62483

Florida Power and Light Company, et 

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

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

County, Florida 

Date of application for amendments:

October 29, 2003. 

Brief description of amendments:

These amendments relocate specific 

pressure and flow values associated 

with the high pressure safety injection, 

low pressure safety injection, boric acid 

makeup, and containment spray pumps 

from the Technical Specification to the 

St. Lucie Units 1 and 2 Updated Final 

Safety Analysis Reports. 

Date of Issuance: October 6, 2004. 

Effective Date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: 194, 136. 

Renewed Facility Operating License 

Nos. DPR–67 and NPF–16: Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: January 6, 2004 (69 FR 697).

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 6, 2004. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket No. 50–305, Kewaunee Nuclear 

Power Plant, Kewaunee County, 

Wisconsin 

Date of application for amendment:

July 6, 2004. 

Brief description of amendment: The 

amendment revises technical 

specification 3.3.a.2.B, by extending the 

completion time from 1 hour to 24 

hours for an accumulator that is 

inoperable for any reason other than 

failure to meet minimum boron 

concentration requirements. 

Date of issuance: October 5, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 178. 

Facility Operating License No. DPR–

43: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 31, 2004 (69 FR 

53111). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 5, 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:

November 1, 2002, and its supplement 

dated April 2, 2004. 

Brief description of amendments: The 

amendments (1) change the allowances 

for bypassing and tripping tested 

channels, (2) remove a surveillance 

requirement for reactor trip system 

(RTS) turbine trip-turbine stop valve 

closure, (3) revise the nominal trip 

setpoint for RTS turbine trip-turbine 

stop valve closure, (4) revise the 

allowable value and nominal trip 

setpoint for RTS interlock, (5) and 

remove and relocate the turbine trip 

function from engineered safety feature 

actuation system turbine trip and 

feedwater isolation to other licenseecontrolled documents. 

Date of issuance: September 24, 2004. 

Effective date: September 24, 2004, 

and shall be implemented within 120 

days from the date of issuance. 

Amendment Nos.: Unit 1–173; Unit 

2–175. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: January 7, 2003 (68 FR 810). 

The April 2, 2004, supplemental letter 

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 September 24, 

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:

May 29, 2003, as supplemented by letter 

dated December 23, 2003, and May 7, 

2004. 

Brief description of amendments: The 

amendments revise several surveillance 

requirements (SRs) in Technical 

Specification (TS) 3.8.1 on alternating 

current sources for plant operation. The 

revised SRs have notes deleted or 

modified to adopt in part Staff-approved 

TSTF–283, Revision 3, which will allow 

these revised SRs to be performed, or 

partially performed, in reactor modes 

that previously were not allowed by the 

TSs. The proposed changes to SRs 

3.8.4.7 and 3.8.4.8 for direct current 

sources were withdrawn in the 

licensee’s letter dated May 7, 2004. 

Date of issuance: September 28, 2004. 

Effective date: September 28, 2004, 

and shall be implemented within 60 

days of the date of issuance including 

the incorporation of the changes to the 

Technical Specification Bases for 

Technical Specification 3.8.1 as 

described in the licensee’s letters dated 

May 29 and December 23, 2003, and 

May 7, 2004, and the NRC safety 

evaluation attached to the amendments. 

This includes the revision of procedures 

to instruct operator action to be taken to 

manually reset the emergency diesel 

generator, as discussed in Section 4.3 of 

the licensee’s May 29, 2003, letter. 

Amendment Nos.: Unit 1–174; Unit 

2–176. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: July 8, 2003 (68 FR 40715). 

The December 23, 2003, and May 7, 

2004, supplemental letters provided 

additional clarifying information, 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 September 28, 

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:

October 22, 2003. 

Brief description of amendments: The 

amendments revise Section 3.6.3 of the 

Diablo Canyon Power Plant Technical 

Specifications to extend the local 

leakage rate testing intervals for the 

containment purge and vent valves with 

resilient seals from 184 days to 24 

months. 

Date of issuance: October 6, 2004. 

Effective date: October 6, 2004, and 

shall be implemented within 60 days 

from the date of issuance. 

Amendment Nos.: Unit 1–175; Unit 

2–177. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: November 25, 2003 (68 FR 

66139). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated October 6, 2004. 

No significant hazards consideration 

comments received: No. 

62484 Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 

Pacific Gas and Electric Company, 

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

Canyon Nuclear Power Plant, Unit Nos. 

1 and 2, San Luis Obispo County, 

California 

Date of application for amendments:

December 19, 2003, and its supplement 

dated May 13, 2004. 

Brief description of amendments: The 

amendments change Technical 

Specification 5.5.9, ‘‘Steam Generator 

(SG) Tube Surveillance Program,’’ to 

revise the wedge region exclusion zones 

for outside diameter stress corrosion 

cracking alternate repair criteria (ARC) 

at tube support plate (TSP) intersections 

and for primary water stress corrosion 

cracking ARC at dented TSP 

intersections. The new wedge region 

exclusion zones are based on new 

analyses of loss-of-coolant accident plus 

safe shutdown earthquake loads 

completed in 2003 using plant-specific 

accident loads. 

Date of issuance: October 6, 2004. 

Effective date: October 6, 2004, and 

shall be implemented within 60 days of 

issuance. 

Amendment Nos.: Unit 1–176; Unit 

2–178. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: February 3, 2004 (69 FR 

5205). 

The May 13, 2004, supplemental 

letter 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 6, 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:

November 17, 2003, as supplemented 

July 15, and August 23, 2004. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications (TSs) to delete the 

primary containment isolation valves 

and instrumentation associated with the 

permanent removal of the reactor vessel 

head spray piping. 

Date of issuance: October 5, 2004. 

Effective date: As of the date of 

issuance, to be implemented prior to 

restart from the fall 2004 refueling 

outage. 

Amendment No.: 152. 

Facility Operating License No. NPF–

57: This amendment revised the TSs. 

Date of initial notice in Federal 

Register: January 20, 2004 (69 FR 

2746). The supplements dated July 15, 

and August 23, 2004, contained 

clarifying information and did not 

change the staff’s proposed finding of no 

significant hazards consideration. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 5, 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:

October 24, 2003, as supplemented by 

letter dated June 29, 2004. 

Brief description of amendment: The 

amendment revised the Surveillance 

Requirements (SRs) associated with 

reactor protection system 

instrumentation. Specifically, the 

amendment revised the SRs associated 

with the control rod block 

instrumentation, source range monitors, 

and power distribution limits by 

removing unnecessary testing 

requirements. 

Date of issuance: October 13, 2004. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 153.

Facility Operating License No. NPF–

57: This amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: December 9, 2003 (68 FR 

68672). The June 29, 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 October 13, 

2004. 

No significant hazards consideration 

comments received: No. 

R.E. Ginna Nuclear Power Plant, LLC, 

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

Power Plant, Wayne County, New York 

Date of application for amendment:

March 1, 2004. 

Brief description of amendment: The 

amendment extends the completion 

time (CT) from 1 hour to 24 hours for 

Condition B of Technical Specification 

(TS) 3.5.1, which defines requirements 

for the emergency core cooling system 

accumulators. Condition B of TS 3.5.1 

specifies a CT to restore an accumulator 

to operable status when it has been 

declared inoperable for a reason other 

than the boron concentration of the 

water in the accumulator not being 

within the required range. 

Date of issuance: October 4, 2004. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment No.: 86. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: June 22, 2004 (69 FR 34706). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 4, 2004. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348, Joseph M. 

Farley Nuclear Plant, Unit 1, Houston 

County, Alabama 

Date of amendments request:

September 19, 2003, as supplemented 

by letters dated March 31, June 18, and 

August 6, 2004. 

Brief Description of amendments:

This amendment revised Technical 

Specifications (TS) Limiting Conditions 

for Operation 3.8.4, ‘‘DC Sources—

Operating,’’ for the remainder of 

operating cycle 19. Specifically, the TS 

change increased the Completion Time 

for the 1B Auxiliary Building DC 

electrical power system inoperability 

due to an inoperable battery to allow for 

on-line replacement of individual cells. 

Date of issuance: September 30, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 164. 

Facility Operating License Nos. NPF–

2: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: November 12, 2003 (68 FR 

64137). 

The supplements dated March 31, 

June 18 and August 6, 2004, provided 

clarifying information that did not 

change the scope of the September 19, 

2003, 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 September 29, 

2004. 

No significant hazards consideration 

comments received: No.

Federal Register / Vol. 69, No. 206 / Tuesday, October 26, 2004 / Notices 62485

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of amendments request: August 

29, 2003, as supplemented by letters 

dated November 11, 2003, and May 5, 

June 10, August 5, August 25, and 

September 27, 2004. 

Brief Description of amendments: The 

amendments revised Technical 

Specifications Limiting Condition of 

Operation 3.9.3, ‘‘Containment 

Penetrations.’’ The changes allow the 

equipment hatch to be open during core 

alterations and/or during movement of 

irradiated fuel assemblies within 

containment. 

Date of issuance: September 30, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 165 and 157. 

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the 

Technical Specifications.

Date of initial notice in Federal 

Register: November 12, 2003 (68 FR 

64137). 

The supplements dated November 11, 

2003, and May 5, June 10, August 5, 

August 25, and September 27, 2004, 

provided clarifying information that did 

not change the scope of the August 29, 

2003, 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 September 30, 

2004. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating 

Company, Inc., Docket Nos. 50–348 and 

50–364, Joseph M. Farley Nuclear Plant, 

Units 1 and 2, Houston County, 

Alabama 

Date of amendments request: August 

25, 2004, as supplemented by letter 

dated September 27, 2004. 

Brief Description of amendments: The 

amendments address the control room 

habitability guidance of Regulatory 

Guide 1.196, ‘‘Control Room 

Habitability at Light-Water Nuclear 

Power Reactors,’’ by revising Limiting 

Condition for Operation 3.7.10, ‘‘Control 

Room Emergency Filtration/

Pressurization System (CREFS)’’ and TS 

5.5.11, ‘‘Ventilation Filter Testing 

Program. The amendments also add a 

new section, TS 5.5.18, ‘‘Control Room 

Integrity Program (CRIP).’’

Date of issuance: September 30, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 166, 158. 

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the 

Technical Specifications. 

Date of initial notice in Federal 

Register: August 31, 2004 (69 FR 

53095). The supplement dated 

September 27, 2004, provided clarifying 

information that did not change the 

scope of the August 25, 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 September 30, 

2004. 

No significant hazards consideration 

comments received: No. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri

Date of application for amendment:

April 8, 2004, as supplemented by letter 

dated September 24, 2004. 

Brief description of amendment: The 

amendment revises requirements in the 

technical specifications to adopt the 

provisions of Industry/Technical 

Specification Task Force (TSTF) change 

TSTF–359, ‘‘Increase Flexibility in 

Mode Restraints.’’ The availability of 

TSTF–359 for adoption by licensees was 

announced in the Federal Register on 

April 4, 2003 (68 FR 16579). 

Date of issuance: October 8, 2004. 

Effective date: October 8, 2004, and 

shall be implemented within 90 days of 

the date of issuance. 

Amendment No.: 164. 

Facility Operating License No. NPF–

30: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: May 11, 2004 (69 FR 26194). 

The additional information provided 

in the supplemental letter dated 

September 24, 2004, does not expand 

the scope of the application as noticed 

and does not change the NRC staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 8, 2004. 

No significant hazards consideration 

comments received: No. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: February 

9, 2004, as supplemented by the letter 

dated September 14, 2004. 

Brief description of amendment: The 

amendment revises requirements in the 

technical specifications to adopt the 

provisions of Industry/Technical 

Specification Task Force (TSTF) change 

TSTF–359, ‘‘Increase Flexibility in 

Mode Restraints.’’

Date of issuance: October 7, 2004. 

Effective date: October 7, 2004, and 

shall be implemented within 90 days of 

the date of issuance. 

Amendment No.: 155. 

Facility Operating License No. NPF–

42. The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: March 16, 2004 (69 FR 

12373). 

The additional information provided 

in the supplemental letter dated 

September 14, 2004, does not expand 

the scope of the application as noticed 

and does not change the NRC staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 7, 2004.

No significant hazards consideration 

comments received: No.

Dated in Rockville, Maryland, this 18th 

day of October 2004.

For the Nuclear Regulatory Commission. 

Ledyard B. Marsh, 

Director, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation.

[FR Doc. 04–23664 Filed 10–25–04; 8:45 am] 

BILLING CODE 7590–01–P

NUCLEAR REGULATORY 

COMMISSION 

[NUREG–1600] 

Revision of the NRC Enforcement 

Policy

AGENCY: Nuclear Regulatory 

Commission.

ACTION: Policy Statement: revision.

SUMMARY: The Nuclear Regulatory 

Commission (NRC or Commission) is 

publishing a revision to its General 

Statement of Policy and Procedure for 

NRC Enforcement Actions (NUREG–

1600) (Enforcement Policy or Policy) to 

address the requirements of the Federal 

Civil Penalties Inflation Adjustment Act 

of 1990, as amended by the Debt 

Collection Improvement Act of 1996. 

The Act requires Federal agencies to 

adjust civil monetary penalties to reflect 

inflation.

DATES: This action is effective on 

November 26, 2004. Comments on this 

revision should be submitted on or 

before December 27, 2004, and will be 

considered by the NRC before the next