Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 60677

TABLE 2.—SUMMARY OF RADIOLOGICAL ENVIRONMENTAL IMPACTS

Radiological Waste Stream ......... No change in design or operation of waste streams. 

Gaseous Waste ........................... Slight increase in amount of radioactive material in gaseous effluents; within FES estimate; offsite doses 

would continue to be well within NRC criteria. 

Liquid Waste ................................ Slight increase in amount of radioactive material in liquid effluents; within FES estimate; offsite doses would 

continue to be well within NRC criteria. 

Solid Waste ................................. No significant change in radioactive resins; no significant changes in dry waste; no significant changes in irradiated components. 

Dose Impacts Occupational Dose Up to 9.6 percent increase in collective occupational dose possible; well within FES estimate. 

Offsite Direct Dose ...................... Slight increase possible; not significant; offsite doses would continue to be within NRC criteria. 

Postulated Accidents ................... Up to 9.6 percent increase in calculated doses from some postulated accidents; calculated doses within NRC 

criteria. 

Fuel Cycle and Transportation .... Increase in bundle average enrichment. Fuel enrichment and burnup would continue to be within bounding 

assumptions for Tables S–3 and S–4 in 10 CFR Part 51, ‘‘Environmental Protection Regulations for Domestic Licensing and Related Regulatory Function;’’ conclusions of tables regarding impact would remain 

valid. 

Alternatives to Proposed Action 

As an alternative to the proposed 

action, the NRC staff considered denial 

of the proposed EPU (i.e., the ‘‘no-action 

alternative’’). Denial of the application 

would result in no change in the current 

environmental impacts; however, other 

fossil-fuel generating facilities may need 

to be built in order to maintain 

sufficient power-generating capacity. As 

an alternative, the licensee could 

purchase power from power generating 

facilities outside the service area. The 

additional power would likely also be 

generated by fossil fuel facilities. 

Construction and operation of a fossilfueled plant would create impacts in air 

quality, land use, and waste 

management significantly greater than 

those identified for the EPU at 

Waterford 3. 

Implementation of the proposed EPU 

would have less impact on the 

environment than the construction and 

operation of a new fossil-fueled 

generating facility or the operator of 

fossil facilities outside the service area. 

Furthermore, the EPU does not involve 

environmental impacts that are 

significantly different from those 

presented in the 1981 FES for Waterford 

3. 

Alternative Use of Resources 

This action does not involve the use 

of any resources not previously 

considered in the 1981 FES for 

Waterford 3. 

Agencies and Persons Consulted 

In accordance with its stated policy, 

on August 13, 2004, the NRC staff 

consulted with the Louisiana State 

official, Ms. Nan Calhoun of the LDEQ, 

regarding the environmental impact of 

the proposed action. The State official 

had no comments. 

Finding of No Significant Impact 

On the basis of the environmental 

assessment, the NRC concludes that the 

proposed action will not have a 

significant effect on the quality of the 

human environment. Accordingly, the 

NRC has determined not to prepare an 

environmental impact statement for the 

proposed action.

DATES: The comment period expires 

November 12, 2004. Comments received 

after this date will be considered if it is 

practical to do so, but the Commission 

is able to assure consideration of 

comments received on or before this 

date.

ADDRESSES: Submit written comments 

to Chief, Rules and Directives Branch, 

Office of Administration, U.S. Nuclear 

Regulatory Commission, Mail Stop T–6 

D59, Washington, DC 20555–0001. 

Written comments may also be 

delivered to 11545 Rockville Pike, Room 

T–6D59, Rockville, Maryland 20852, 

from 7:30 a.m. to 4:15 p.m. on Federal 

workdays. Copies of written comments 

received will be electronically available 

at the NRC’s Public Electronic Reading 

Room link http://www.nrc.gov/readingrm/adams.html on the NRC Homepage 

or at the NRC’s Public Document Room 

located at One White Flint North, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. 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 Reference staff at 1–

800–397–4209, or 301–415–4737, or by 

e-mail at pdr@nrc.gov.

SUPPLEMENTARY INFORMATION: The NRC 

is considering issuance of an 

amendment to Facility Operating 

License No. NPF–38 issued to Entergy 

for operation of Waterford 3 located in 

St. Charles Parish, Louisiana.

FOR FURTHER INFORMATION CONTACT: N. 

Kalyanam, Office of Nuclear Reactor 

Regulation, Mail Stop O–7D1, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, by 

telephone at (301) 415–1480, or by email at nxk@nrc.gov.

Dated at Rockville, Maryland, this 30th day 

of September 2004.

For the Nuclear Regulatory Commission. 

Michael K. Webb, 

Acting Chief, Section 1, Project Directorate 

IV, Division of Licensing Project Management, 

Office of Nuclear Reactor Regulation.

[FR Doc. 04–22786 Filed 10–8–04; 8:45 am] 

BILLING CODE 7590–01–P

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued, from September 

17, 2004, through September 30, 2004. 

The last biweekly notice was published 

on September 28, 2004 (69 FR 57978). 

60678 Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. Within 60 days after the 

date of publication of this notice, the 

licensee may file a request for a hearing 

with respect to issuance of the 

amendment to the subject facility 

operating license and any person whose 

interest may be affected by this 

proceeding and who wishes to 

participate as a party in the proceeding 

must file a written request for a hearing 

and a petition for leave to intervene. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60-

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules and 

Directives Branch, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be delivered to Room 6D22, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

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

Copies of written comments received 

may be examined at the Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. The filing of 

requests for a hearing and petitions for 

leave to intervene is discussed below. 

Within 60 days after the date of 

publication of this notice, the licensee 

may file a request for a hearing with 

respect to issuance of the amendment to 

the subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

intervene. Requests for a hearing and a 

petition for leave to intervene shall be 

filed in accordance with the 

Commission’s ‘‘Rules of Practice for 

Domestic Licensing Proceedings’’ in 10 

CFR Part 2. Interested persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/

requestor to relief. A petitioner/

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party.

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 60679

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 email to 

pdr@nrc.gov. 

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 make 

changes to the Technical Specifications 

(TSs) that will eliminate secondary 

containment operability requirements 

when handling sufficiently decayed 

irradiated fuel and performing core 

alterations, and will clarify 

requirements associated with operations 

with potential to drain the reactor 

vessel. This proposed amendment also 

uses Alternate Source Term (AST) 

methodology in accordance with 10 CFR 

50.67 for calculating Fuel Handling 

Accident (FHA) consequences. The 

proposed amendment also removes TSs 

operability requirements for engineered 

safety features (ESF) (e.g. primary/

secondary containment, standby gas 

treatment, and isolation capability) after 

the sufficient decay of ‘‘recently’’

irradiated fuel. 

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. Do the Proposed Changes Involve a 

Significant Increase in the Probability or 

Consequence of an Accident Previously 

Evaluated? 

The proposed changes do not modify 

the design or operation of equipment 

used to handle and move new and spent 

fuel or to perform core alterations. The 

proposed amendment does not modify 

the design of the ESF equipment. The 

proposed changes, therefore, will not 

increase the probability of accidents 

previously evaluated. 

AST analysis does not affect the 

performance of the systems or 

components used to mitigate the 

consequences of accidents previously 

evaluated. While a direct comparison 

between current methodologies used in 

the current Pilgrim design basis analysis 

and Regulatory Guide (RG) 1.183 is not 

possible due to different acceptance 

criteria, the AST calculations 

demonstrate that the radiological 

consequences to the accidents 

previously evaluated will still remain 

below the regulatory limits. Therefore, 

any potential change in the radiological 

consequences are not considered 

significant. 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. Do the Proposed Changes Create the 

Possibility of a New or Different Kind of 

Accident From Any Accident 

Previously Analyzed? 

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. Do the Proposed Changes Involve a 

Significant Reduction in the Margin of 

Safety? 

The licensee performed a 

comprehensive analysis and evaluation 

of the FHA using AST methodology and 

dose consequence analysis in 

accordance with 10 CFR 50.67. While 

direct comparison between 

methodologies used in the current 

Pilgrim design basis analysis and RG 

1.183 is not possible due to different 

acceptance criteria, the revised doses 

will, however, remain below the total 

effective dose equivalent dose 

regulatory limits for the control room, 

exclusion area boundary, and low 

population zone as specified in 10 CFR 

50.67. Therefore, by meeting the 

applicatory regulatory limits for AST, 

any potential decrease in a margin of 

safety would not be considered 

significant. The changes are, therefore, 

not considered a significant 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: J. M. Fulton, 

Esquire, Assistant General Counsel, 

Pilgrim Nuclear Power Station, 600 

Rocky Hill Road, Plymouth, 

Massachusetts, 02360–5599. 

NRC Acting Section Chief: Daniel S. 

Collins.

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request: June 17, 

2004. 

Description of amendment request: 

The proposed amendment would delete 

60680 Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 

entries from Technical Specification 

(TS) Tables 3.2.6 and 4.2.6 related to the 

post-accident hydrogen and oxygen 

monitors. Licensees were generally 

required to implement upgrades as 

described in NUREG–0737, 

‘‘Clarification of TMI [Three Mile 

Island] Action Plan Requirements,’’ and 

Regulatory Guide (RG) 1.97, 

‘‘Instrumentation for Light-WaterCooled Nuclear Power Plants to Assess 

Plant and Environs Conditions During 

and Following an Accident.’’

Implementation of these upgrades was 

an outcome of the lessons learned from 

the accident that occurred at TMI, Unit 

2. Requirements related to combustible 

gas control were imposed by Order for 

many facilities and were added to or 

included in the TSs for nuclear power 

reactors currently licensed to operate. 

The revised 10 CFR 50.44, ‘‘Combustible 

gas control system for nuclear power 

reactors,’’ eliminated the requirements 

for hydrogen recombiners (not installed 

at Vermont Yankee and therefore not 

addressed by this proposed amendment) 

and relaxed safety classifications and 

licensee commitments to certain design 

and qualification criteria for hydrogen 

and oxygen monitors. 

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 September 25, 2003 (68 FR 

55416). The licensee affirmed the 

applicability of the model NSHC 

determination in its application dated 

June 17, 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 revised 10 CFR 50.44 no longer defines 

a design-basis loss-of-coolant accident 

(LOCA) hydrogen release, and eliminates 

requirements for hydrogen control systems to 

mitigate such a release. The installation of 

hydrogen recombiners and/or vent and purge 

systems required by 10 CFR 50.44(b)(3) was 

intended to address the limited quantity and 

rate of hydrogen generation that was 

postulated from a design-basis LOCA. The 

Commission has found that this hydrogen 

release is not risk-significant because the 

design-basis LOCA hydrogen release does not 

contribute to the conditional probability of a 

large release up to approximately 24 hours 

after the onset of core damage. In addition, 

these systems were ineffective at mitigating 

hydrogen releases from risk-significant 

accident sequences that could threaten 

containment integrity. 

With the elimination of the design-basis 

LOCA hydrogen release, hydrogen and 

oxygen monitors are no longer required to 

mitigate design-basis accidents and, 

therefore, the hydrogen monitors do not meet 

the definition of a safety-related component 

as defined in 10 CFR 50.2. RG 1.97 Category 

1, is intended for key variables that most 

directly indicate the accomplishment of a 

safety function for design-basis accident 

events. The hydrogen and oxygen monitors 

no longer meet the definition of Category 1 

in RG 1.97. As part of the rulemaking to 

revise 10 CFR 50.44, the Commission found 

that Category 3, as defined in RG 1.97, is an 

appropriate categorization for the hydrogen 

monitors because the monitors are required 

to diagnose the course of beyond design-basis 

accidents. Also, as part of the rulemaking to 

revise 10 CFR 50.44, the Commission found 

that Category 2, as defined in RG 1.97, is an 

appropriate categorization for the oxygen 

monitors, because the monitors are required 

to verify the status of the inert containment. 

The regulatory requirements for the 

hydrogen and oxygen monitors can be 

relaxed without degrading the plant 

emergency response. The emergency 

response, in this sense, refers to the 

methodologies used in ascertaining the 

condition of the reactor core, mitigating the 

consequences of an accident, assessing and 

projecting offsite releases of radioactivity, 

and establishing protective action 

recommendations to be communicated to 

offsite authorities. Classification of the 

hydrogen monitors as Category 3, 

classification of the oxygen monitors as 

Category 2, and removal of the hydrogen and 

oxygen monitors from TSs will not prevent 

an accident management strategy through the 

use of the severe accident management 

guidelines, the emergency plan, the 

emergency operating procedures, and site 

survey monitoring that support modification 

of emergency plan protective action 

recommendations. 

Therefore, the relaxation of the hydrogen 

and oxygen monitor requirements, including 

removal of these requirements from TSs, does 

not involve a significant increase in the 

probability or the consequences of any 

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 relaxation of the hydrogen and oxygen 

monitor requirements, including removal of 

these requirements from TSs, will not result 

in any failure mode not previously analyzed. 

The hydrogen and oxygen monitor 

equipment was intended to mitigate a designbasis hydrogen release. The hydrogen and 

oxygen monitor equipment are not 

considered accident precursors, nor does 

their existence or elimination have any 

adverse impact on the pre-accident state of 

the reactor core or post accident confinement 

of radionuclides within the containment 

building. 

Therefore, 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 relaxation of the hydrogen and oxygen 

monitor requirements, including removal of 

these requirements from TSs, in light of 

existing plant equipment, instrumentation, 

procedures, and programs that provide 

effective mitigation of and recovery from 

reactor accidents, results in a neutral impact 

to the margin of safety. 

The installation of hydrogen recombiners 

and/or vent and purge systems required by 

10 CFR 50.44(b)(3) was intended to address 

the limited quantity and rate of hydrogen 

generation that was postulated from a designbasis LOCA. The Commission has found that 

this hydrogen release is not risk-significant 

because the design-basis LOCA hydrogen 

release does not contribute to the conditional 

probability of a large release up to 

approximately 24 hours after the onset of 

core damage. 

Category 3 hydrogen monitors are adequate 

to provide rapid assessment of current 

reactor core conditions and the direction of 

degradation while effectively responding to 

the event in order to mitigate the 

consequences of the accident. The intent of 

the requirements established as a result of the 

TMI, Unit 2 accident can be adequately met 

without reliance on safety-related hydrogen 

monitors. 

Category 2 Oxygen Monitors Are Adequate 

To Verify the Status of an Inerted 

Containment. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

The intent of the requirements established as 

a result of the TMI, Unit 2 accident can be 

adequately met without reliance on safetyrelated oxygen monitors. Removal of 

hydrogen and oxygen monitoring from TSs 

will not result in a significant reduction in 

their functionality, reliability, and 

availability.

Based on the reasoning presented 

above and the previous discussion of 

the amendment request, the requested 

change does not involve a significant 

hazards consideration. 

Attorney for licensee: Mr. David R. 

Lewis, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037–1128. 

NRC Section Chief: Allen G. Howe. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request: 

September 1, 2004. 

Description of amendment request: 

The proposed amendment would delete 

Technical Specification (TS) 6.6.A, 

‘‘Occupational Radiation Exposure 

Report,’’ and TS 6.6.B, ‘‘Monthly 

Operating Reports.’’

The NRC staff issued a notice of 

availability of a model no significant 

Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 60681

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 1, 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? 

Response: No. 

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.

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration of the plant, add any new 

equipment, or require any existing 

equipment to be operated in a manner 

different from the present design. Therefore, 

the proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

This is an administrative change to 

reporting requirements of plant operating 

information and occupational radiation 

exposure data, and has no effect on plant 

equipment, operating practices or safety 

analyses assumptions. For these reasons, the 

proposed change does not involve a 

significant reduction in the margin of safety.

Based upon the reasoning presented 

above, the requested change does not 

involve a significant hazards 

consideration. 

Attorney for licensee: Mr. David R. 

Lewis, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037–1128. 

NRC Section Chief: Allen G. Howe. 

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 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 TSs 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 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 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 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 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.

60682 Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 

FirstEnergy Nuclear Operating 

Company, Docket No. 50–346, DavisBesse Nuclear Power Station, Unit 1, 

Ottawa County, Ohio 

Date of amendment request: August 2, 

2004. 

Description of amendment request:

The proposed amendment would delete 

Technical Specification (TS) Section 

6.8.4, ‘‘Post Accident Sampling,’’ and 

the related requirements to maintain a 

Post-Accident Sampling System (PASS). 

Licensees were generally required to 

implement PASS upgrades as described 

in NUREG–0737, ‘‘Clarification of TMI 

[Three Mile Island] Action Plan 

Requirements,’’ and Regulatory Guide 

1.97, Revision 3, ‘‘Instrumentation for 

Light-Water-Cooled Nuclear Power 

Plants to Access Plant and Environs 

Conditions During and Following an 

Accident.’’ Implementation of these 

upgrades was an outcome of the U.S. 

Nuclear Regulatory Commission’s (NRC) 

lessons learned from the accident that 

occurred at TMI Unit 2. Requirements 

related to PASS were imposed by Order 

for many facilities and were added to or 

included in the TS for nuclear power 

reactors currently licensed to operate. 

Lessons learned and improvements 

implemented over the last 20 years have 

shown that the information obtained 

from PASS can be readily obtained 

through other means or is of little use 

in the assessment and mitigation of 

accident conditions. 

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on March 3, 2003 (68 FR 

10052) on possible amendments to 

eliminate PASS, 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 a license 

amendment application in the Federal 

Register on May 13, 2003 (68 FR 25664). 

The licensee affirmed the applicability 

of the following NSHC determination in 

its application dated August 2, 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 PASS was originally designed to 

perform many sampling and analysis 

functions. These functions were designed 

and intended to be used in post accident 

situations and were put into place as a result 

of the TMI–2 accident. The specific intent of 

the PASS was to provide a system that has 

the capability to obtain and analyze samples 

of plant fluids containing potentially high 

levels of radioactivity, without exceeding 

plant personnel radiation exposure limits. 

Analytical results of these samples would be 

used largely for verification purposes in 

aiding the plant staff in assessing the extent 

of core damage and subsequent offsite 

radiological dose projections. The system 

was not intended to and does not serve a 

function for preventing accidents and its 

elimination would not affect the probability 

of accidents previously evaluated. 

In the 20 years since the TMI–2 accident 

and the consequential promulgation of post 

accident sampling requirements, operating 

experience has demonstrated that a PASS 

provides little actual benefit to post accident 

mitigation. Past experience has indicated that 

there exists in-plant instrumentation and 

methodologies available in lieu of a PASS for 

collecting and assimilating information 

needed to assess core damage following an 

accident. Furthermore, the implementation of 

Severe Accident Management Guidance 

(SAMG) emphasizes accident management 

strategies based on in-plant instruments. 

These strategies provide guidance to the 

plant staff for mitigation and recovery from 

a severe accident. Based on current severe 

accident management strategies and 

guidelines, it is determined that the PASS 

provides little benefit to the plant staff in 

coping with an accident. 

The regulatory requirements for the PASS 

can be eliminated without degrading the 

plant emergency response. The emergency 

response, in this sense, refers to the 

methodologies used in ascertaining the 

condition of the reactor core, mitigating the 

consequences of an accident, assessing and 

projecting offsite releases of radioactivity, 

and establishing protective action 

recommendations to be communicated to 

offsite authorities. The elimination of the 

PASS will not prevent an accident 

management strategy that meets the initial 

intent of the post-TMI–2 accident guidance 

through the use of the SAMGs, the 

emergency plan (EP), the emergency 

operating procedures (EOP), and site survey 

monitoring that support modification of 

emergency plan protective action 

recommendations (PARs). 

Therefore, the elimination of PASS 

requirements from Technical Specification 

(TS) (and other elements of the licensing 

bases) does not involve a significant increase 

in the consequences of any 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 elimination of PASS related 

requirements will not result in any failure 

mode not previously analyzed. The PASS 

was intended to allow for verification of the 

extent of reactor core damage and also to 

provide an input to offsite dose projection 

calculations. The PASS is not considered an 

accident precursor, nor does its existence or 

elimination have any adverse impact on the 

pre-accident state of the reactor core or post 

accident confinement of radioisotopes within 

the containment building. 

Therefore, 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 elimination of the PASS, in light of 

existing plant equipment, instrumentation, 

procedures, and programs that provide 

effective mitigation of and recovery from 

reactor accidents, results in a neutral impact 

to the margin of safety. Methodologies that 

are not reliant on PASS are designed to 

provide rapid assessment of current reactor 

core conditions and the direction of 

degradation while effectively responding to 

the event in order to mitigate the 

consequences of the accident. The use of a 

PASS is redundant and does not provide 

quick recognition of core events or rapid 

response to events in progress. The intent of 

the requirements established as a result of the 

TMI–2 accident can be adequately met 

without reliance on a PASS. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

Based upon the reasoning presented above 

and the previous discussion of the 

amendment request, the requested change 

does not involve a significant hazards 

consideration.

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mary E. 

O’Reilly, Attorney, FirstEnergy 

Corporation, 76 South Main Street, 

Akron, OH 44308. 

NRC Section Chief: Anthony J. 

Mendiola. 

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: April 

13, 2004.

Description of amendment requests: 

The proposed amendments would 

change the licensing basis as described 

in the Updated Final Safety Analysis 

Report to allow the use of a reinforcing 

bar (rebar) yield strength value based on 

measured material properties, as 

documented in the licensee rebar 

acceptance tests, in control rod drive 

missile shield structural 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. Does the proposed change involve a 

significant increase in the probability of 

occurrence or consequences of an accident 

previously evaluated? 

Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 60683

Response: No 

Probability of Occurrence of an Accident 

Previously Evaluated 

This is a change in the method of 

determining the acceptability of 

accommodating the pressure load following a 

loss-of-coolant accident. No physical changes 

are being made to the plant and no potential 

accident initiators are introduced by this 

change. Thus, the probability of the 

occurrence of any accident previously 

evaluated is not significantly increased. 

Consequences of an Accident Previously 

Evaluated 

There is reasonable assurance that the 

ability of control rod drive missile shields 

(missile shields) to maintain their structural 

capability and continue to function as a part 

of the divider barrier separating the lower 

containment from the upper containment is 

not impacted by this change. The data 

obtained from rebar acceptance test reports 

demonstrate that the missile shields have 

adequate strength to accommodate the load 

that would be imposed under assumed 

accident conditions. As a result, the 

consequences of any accident previously 

evaluated are not significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No 

The use of increased missile shield rebar 

yield strength for the missile shield structural 

capability under accident conditions does 

not alter the evaluation of the missile shields’

structural capability during normal 

operation, the operational condition in which 

a new or different kind of accident would be 

initiated. The change does not physically 

alter plant components nor does it alter plant 

operation. The change does not adversely 

affect current system interfaces or create new 

interfaces that could result in an accident or 

malfunction of a different kind than 

previously evaluated. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No 

The margin of safety for the missile shields 

is provided by the factors that are applied to 

the individual loads determining the load 

imposed on the missile shields under 

accident conditions. These code safety 

factors are sufficient to ensure that both 

anticipated and unanticipated loads can be 

withstood by the concrete structures. The use 

of yield strengths based on measured 

material properties as documented in the 

I&M [Indiana Michigan Power Company] 

rebar acceptance tests for the missile shield 

structural evaluation has no effect on the 

margin of safety provided by the load safety 

factors. I&M continues to use the same load 

factors that were used to license the Donald 

C. Cook Nuclear Plant. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 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 revises Fort 

Calhoun Station (FCS) Technical 

Specification (TS) 5.9.5, ‘‘Core 

Operating Limits Report,’’ such that it 

will read consistent with TS 5.6.5 of 

NUREG–1432, Standard Technical 

Specifications-Combustion Engineering 

Plants. In addition, the list of core 

reload analysis methodologies 

contained in TS 5.9.5b used to 

determine the core operating limits is 

updated to move many of these 

references to Omaha Public Power 

District (OPPD) core reload analysis 

methodology documents OPPD–NA–

8301, 8302, and 8303. Several analytical 

method references that are no longer 

applicable to FCS are deleted from TS 

5.9.5b; several references will remain, as 

they are not suitable for incorporation 

into the core reload analysis documents. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. The proposed changes are 

primarily administrative in nature to achieve 

consistency with Standard Technical 

Specifications and to update the list of NRC 

reviewed and approved analytical methods 

used to develop core operating limits. Several 

of the analytical methods are no longer 

applicable to Fort Calhoun Station, Unit No. 

1, (FCS) and thus are deleted from the 

Technical Specifications (TSs). Many of the 

topical reports currently referenced in TS 

5.9.5b are more suitably referenced in the 

OPPD core reload methodology documents 

where they have been relocated. 

The OPPD core reload methodology 

documents remain referenced in TS 5.9.5b 

and as such are subject to NRC review and 

approval. The relocation of the topical 

reports referenced in TS 5.9.5b to OPPD core 

reload methodology documents is an 

administrative change. In addition to the 

incorporation of references currently found 

in TS 5.9.5b, OPPD core reload methodology 

documents OPPD–NA–8301, 8302, and 8303 

are revised to remove characters designating 

them as proprietary, and approved. This is an 

administrative change, as OPPD no longer 

considers the documents to be proprietary or 

topical reports. OPPD core reload 

methodology documents OPPD–NA–8301, 

8302, and 8303 are enclosed for NRC review 

and approval [attached to the licensee’s 

September 7, 2004, letter] of the changes 

noted above and incorporation of the 

CASMO–4 (C–4) computer code, which is 

described below. 

OPPD is adding the C–4 code to OPPD–

NA–8302, Reload Core Analysis 

Methodology, Neutronics Design Methods 

and Verification and will use the code for 

nuclear design analysis. This will allow the 

use of the C–4 and SIMULATE–3 (S–3) 

methodology to perform all steady-state 

pressurized water reactor (PWR) core physics 

analyses. The probability of occurrence of an 

accident previously evaluated will not be 

increased by the proposed change in the 

particular codes used for physics calculations 

for nuclear design analysis. The results of 

nuclear design analyses are used as inputs to 

the analysis of accidents that are evaluated in 

the Updated Safety Analysis Report (USAR). 

These inputs do not alter the physical 

characteristics or modes of operation of any 

system, structure, or component involved in 

the initiation of an accident. Thus, there is 

no significant increase in the probability of 

an accident previously evaluated as a result 

of this change. 

The consequences of an accident evaluated 

in the USAR are affected by the value of 

inputs to the transient safety analysis. An 

extensive benchmark of C–4/S–3 predictions 

was performed with measured data using a 

variety of fuel designs and operating 

conditions in power reactors and critical 

experiments. The accuracy of C–4/S–3 is 

similar to, and sometimes better than, the 

accuracy of C–3/S–3. Furthermore, there is 

always the potential for the value of the 

nuclear design parameters to change solely as 

a result of the new core reload fuel core 

loading pattern. Regardless of the source of 

a change, an assessment is always made of 

changes to the nuclear design parameters 

with respect to their effects on the 

consequences of accidents previously 

evaluated in the USAR. Refueling is an 

anticipated activity, which is described in 

the USAR. If increased consequences are 

anticipated, compensatory actions are 

implemented to neutralize any expected 

increase in consequences. These 

compensatory actions include, but are not 

limited to, crediting any existing margins in 

the analysis or redefining the operating 

envelope to avoid increased consequences. 

Thus, the nuclear design parameters are 

intermediate results and by themselves will 

not result in an increase in the consequence 

of an accident evaluated in the USAR. 

60684 Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 

Therefore, the use of the C–4/S–3 code 

package, which will perform the same 

functions as the C–3/S–3 codes with similar 

accuracy, does not significantly increase the 

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 does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. The proposed changes are 

primarily administrative in nature. The 

changes achieve consistency with Standard 

Technical Specifications, update the list of 

NRC reviewed and approved analytical 

methods used to develop core operating 

limits by deleting certain analytical methods 

no longer applicable to FCS and relocating 

many of the remainder to OPPD core reload 

analysis methodology documents, and make 

minor administrative changes to OPPD core 

reload analysis documents referenced in TS 

5.9.5b. OPPD intends to utilize the C–4/S–3 

code package for nuclear design analysis. The 

proposed amendment would add the C–4 

code to OPPD core reload analysis 

methodology document OPPD–NA–8302. 

The possibility for a new or different kind 

of accident evaluated previously in the USAR 

will not be created by the proposed 

administrative changes or the change to the 

particular codes used for physics calculations 

for nuclear design analyses. The change 

involves adding the Studsvik C–4 code to 

OPPD core reload analysis methodology 

document OPPD–NA–8302. The C–4 code is 

an update to the C–3 code currently 

approved for use at FCS. The results of 

nuclear design analyses are used as inputs to 

the analysis of accidents that are evaluated in 

the USAR. These inputs do not alter the 

physical characteristics or modes of 

operation of any system, structure or 

component involved in the initiation of an 

accident. Therefore, these administrative 

changes and the addition of the C–4 code, 

which will perform the same functions, as 

the C–3 code with similar accuracy, does not 

increase the possibility of a new or different 

kind of accident from any accident 

previously evaluated.

3. The proposed change does not involve 

a significant reduction in a margin of safety. 

The proposed change does not involve a 

significant reduction in a margin of safety. 

The margin of safety as defined in the basis 

for any technical specification will not be 

reduced nor increased by the proposed 

administrative changes or the change to the 

codes used for physics calculations for 

nuclear design analyses. The changes achieve 

consistency with Standard Technical 

Specifications, update the list of NRC 

approved analytical methods used to develop 

core operating limits by deleting certain 

analytical methods no longer applicable to 

FCS and relocating many of the remainder to 

OPPD core reload analysis methodology 

documents, and make minor administrative 

changes to OPPD core reload analysis 

documents referenced in TS 5.9.5b. 

The change involves the addition of the 

Studsvik C–4 code to OPPD core reload 

analysis methodologies for nuclear design 

analysis. Extensive benchmarking of the C–

4/S–3 computer codes has demonstrated that 

the values of those parameters used in the 

safety analysis are not significantly changed 

relative to the values obtained using the NRC 

approved C–3/S–3 computer codes. For any 

changes in the calculated values that do 

occur, the application of appropriate biases 

and uncertainties ensures that the current 

margin of safety is maintained. Specifically, 

use of these code specific biases and 

uncertainties in safety evaluations continues 

to provide the same statistical assurance that 

the values of the nuclear parameters used in 

the safety analysis are conservative with 

respect to the actual values on at least a 95/

95 probability/confidence basis.

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 R. 

Curtiss, Esq., Winston & Strawn, 1400 L 

Street, NW., Washington, DC 20005–

3502. 

NRC Section Chief: Robert Gramm. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of amendment request: April 15, 

2004, as supplemented August 11, 2004. 

Description of amendment request: 

The amendment request proposes to 

revise the Salem Unit No. 1 Technical 

Specifications (TSs) to reflect the 

addition of the chilled water system to 

provide cooling water to the 

containment fan cooling units (CFCUs). 

The amendment request also proposes 

to revise a non-conservative Action 

Statement for Salem Unit Nos. 1 and 2 

that allows three containment cooling 

fans to be inoperable under certain 

conditions. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the change involve a significant 

increase in the probability of occurrence or 

consequences of an accident previously 

evaluated? 

Response: No. 

Containment cooling fans remove 

containment heat loads under both normal 

and accident conditions. As such, they have 

no impact on the probability of occurrence of 

any previously evaluated accidents, although 

they do function to mitigate accident 

consequences. With regard to accident 

consequences, revised containment response 

analysis has been performed with the 

proposed changes of this license amendment. 

This analysis demonstrates that containment 

pressure and temperature limits continue to 

be met as further described below. 

The addition of the non-safety related 

chilled water system does not represent an 

increase in the consequences of an accident 

since, at the onset of the accident, the chilled 

water supply is automatically isolated on the 

resulting safety injection signal and the safety 

related Service Water System supplies the 

cooling method to remove the containment 

heat loads, as presently analyzed. Analysis 

has been performed to evaluate any potential 

failures that could prevent the Containment 

Cooling System to perform [sic] its safety 

related functions. Redundancy in the chilled 

water system and transfer to service water 

during an accident are incorporated in the 

design. In addition, as a conservative 

measure, an action statement has been added 

to require prompt action to restore 

containment cooling or commence a unit 

shutdown in the event of an unexpected 

condition that results in the loss of normal 

containment cooling capability. 

The accidents previously evaluated that are 

associated with containment heat removal 

are design basis loss-of-coolant accident 

(LOCA) and main steam line break (MSLB) 

accident. In the case of the design basis 

LOCA, the revised analysis demonstrates that 

all cases resulted in a peak containment 

pressure that was less than 47 psig. In 

addition, all long-term cases were well below 

50% of the peak value within 24 hours. 

Based on the results, applicable criteria for 

Salem Unit 1 have been met and therefore, 

the consequences of previously evaluated 

accidents are not increased. 

The proposed change to the nonconservative TS 3.6.2.3 Action b, maintains 

that five CFCUs remain operable to ensure 

that, upon a single failure, a minimum of 

three CFCUs will provide the required 

containment and air mixing which is 

consistent with the current Salem Dose 

Analysis. 

Consequently, the proposed license 

amendment does not increase the probability 

of occurrence or the consequences of 

accidents previously evaluated for Salem. 

2. Create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

Response: No. 

Containment cooling fans remove 

containment heat loads under both normal 

and accident conditions. The containment 

cooling fans are presently part of the plant 

protection equipment and have been 

analyzed and evaluated as to their function 

and effectiveness. Consequently, they cannot 

create the possibility of any new or different 

kinds of accidents from any previously 

evaluated. The addition of a chilled water 

system that is isolated on an accident 

condition does not create a new or different 

kind of accident. The accidents analyzed are 

the LOCA and MSLB, which are part of the 

Salem Design Bases. 

The proposed change to the nonconservative TS 3.6.2.3 Action b, maintains 

that five CFCUs remain operable to ensure 

that, upon a single failure, a minimum of 

Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 60685

three CFCUs will provide the required 

containment and air mixing which is 

consistent with the current Salem Dose 

Analysis. 

Therefore, the proposed license 

amendment does not create the possibility of 

a new or different kind of accident from any 

accident previously evaluated. 

3. Does the change involve a significant 

reduction in a margin of safety? 

Response: No. 

The margin of safety pertinent to the 

proposed changes is the dose consequences 

resulting from a design basis LOCA. 

Containment cooling fans affect potential 

dose consequences in that they assist in 

maintaining containment pressure and 

temperature within design limits. By 

maintaining these limits, three critical 

functions are performed. These are: 

a. Containment integrity is assured by 

maintaining pressure below the containment 

design limit. 

b. By maintaining pressure below 47 psig, 

leakage of containment atmosphere to the 

surrounding environment is retained within 

the leakage testing results of 10 CFR 50, 

Appendix J. In this case, the Appendix J 

testing procedures provide the margin of 

safety, as long as the limiting pressure (47 

psig) is not exceeded. 

c. By maintaining containment temperature 

within limits, the qualification of vital 

electrical equipment to function in the postaccident containment environment is 

assured. In this case, the margin of safety is 

provided by the testing and evaluation 

procedures implemented by 10 CFR 50.49. 

In addition, as a conservative measure, an 

action statement has been added to require 

prompt action to restore containment cooling 

or commence a unit shutdown in the event 

of an unexpected condition that results in the 

loss of normal containment cooling 

capability. 

The proposed change to the nonconservative TS 3.6.2.3 Action b, maintains 

that five CFCUs remain operable to ensure 

that, upon a single failure, a minimum of 

three CFCUs will provide the required 

containment and air mixing which is 

consistent with the current Salem Dose 

Analysis.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jeffrie J. Keenan, 

Esquire, Nuclear Business Unit–N21, 

P.O. Box 236, Hancocks Bridge, NJ 

08038. 

NRC Section Chief: James W. Clifford. 

R.E. Ginna Nuclear Power Plant, LLC, 

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

Power Plant, Wayne County, New York 

Date of amendment request: July 26, 

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 July 26, 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 Technical 

Specification 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 NRC staff proposes to 

determine that the requested change 

does not involve significance hazards 

consideration. 

Attorney for licensee: Daniel F. 

Stenger, Ballard Spahr Andrews & 

Ingersoll, LLP, 601 13th Street, NW., 

Suite 1000 South, Washington, DC 

20005. 

NRC Section Chief: Richard J. Laufer. 

South Carolina Electric & Gas Company, 

South Carolina Public Service 

Authority, Docket No. 50–395, Virgil C. 

Summer Nuclear Station (VCSNS), Unit 

No. 1, Fairfield County, South Carolina 

Date of amendment request: August 

24, 2004. 

Description of amendment request: 

The proposed change will revise 

Surveillance Requirements (SRs) 4.7. 

1.2.a.1 and 4.7 .1.2.a.2 to reflect a more 

representative model of the Emergency 

Feedwater (EFW) System. The new 

model has established new technical 

specification (TS) acceptance criteria to 

assure the design requirements of the 

system are met. These required 

characteristics are more stringent than 

those currently in the VCSNS TSs for 

this system. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

This change represents a more restrictive 

surveillance requirement than currently 

exists for TS Surveillance 4.7. 1.2.a.1 and 4.7 

.1.2.a.2. These proposed surveillance 

acceptance criteria changes will ensure that 

the motor driven EFW pumps and the turbine 

driven EFW pump can continue to perform 

their design function. There are no changes 

planned to any plant installed hardware or 

software and normal plant operations will 

not be impacted. 

The probability or consequences of 

accidents previously evaluated in the VCSNS 

FSAR [Final Safety Analysis Report] are 

unaffected by this proposed change because 

there is no change to any equipment response 

or accident mitigation scenario. There are no 

additional challenges to fission product 

barrier integrity. 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 involves the 

revision of the Surveillance Requirements for 

the EFW system. The revised requirements 

are more restrictive to insure compliance 

with the design basis of the system. Changes 

to the system model require changes to the 

SR acceptance criteria in order to maintain 

the performance level assumed in the safety 

analysis. 

60686 Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

change. The proposed change does not 

challenge the performance or integrity of any 

safety-related system. Therefore, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Does this change involve a significant 

reduction in margin of Safety? 

The proposed change will have no affect 

on the availability, operability, or 

performance of the safety-related systems and 

components. A change to the SR is proposed, 

however, the proposed change is more 

restrictive than the current SR. The more 

restrictive criteria inherently include a 5 gpm 

leak tolerance for the EFW flow control 

valves. This represents a built in margin for 

the pump head requirement when the flow 

control valve leakage is determined to be less 

than 5 gpm. Therefore, the proposed change 

does not involve a significant reduction in 

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 G. 

Eppink, South Carolina Electric & Gas 

Company, Post Office Box 764, 

Columbia, South Carolina 29218. 

NRC Section Chief: Mary Jane RossLee, Acting. 

Southern Nuclear Operating Company, 

Inc., et al., Docket Nos. 50–321 and 50–

366, Edwin I. Hatch Nuclear Plant, 

Units 1 and 2, Appling County, Georgia; 

Docket Nos. 50–348 and 50–364, Joseph 

M. Farley Nuclear Plant, Units 1 and 2, 

Houston County, Alabama; Docket Nos. 

50–424 and 50–425, Vogtle Electric 

Generating Plant, Units 1 and 2, Burke 

County, Georgia 

Date of amendment request: July 28, 

2004. 

Description of amendment request: 

The proposed amendment would delete 

Technical Specification (TS) 5.6.1, 

‘‘Occupational Radiation Exposure 

Report,’’ and TS 5.6.4, ‘‘Monthly 

Operating Reports.’’

The NRC staff issued a notice of 

availability of a model no significant 

hazards consideration (NSHC) 

determination for referencing in license 

amendment applications in the Federal 

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

The licensee affirmed the applicability 

of the model NSHC determination in its 

application dated July 28, 2004. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

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 a margin of safety.

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ernest L. Blake, 

Jr., Esquire, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037. 

NRC Section Chief: Mary Jane RossLee, Acting.

Southern Nuclear Operating Company 

(SNC), Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of Georgia, 

City of Dalton, Georgia, Docket Nos. 50–

321 and 50–366, Edwin I. Hatch Nuclear 

Plant, Units 1 and 2, Appling County, 

Georgia 

Date of amendment request: July 20, 

2004. 

Description of amendment request:

The proposed amendments would make 

various changes to the technical 

specifications (TSs) associated with the 

Plant Hatch DC electrical system 

consistent with Technical Specification 

Task Force (TSTF) change traveler 

TSTF–360, including specific action 

and increased completion time for an 

inoperable battery charger, increase the 

completion time for an inoperable 

station service battery from 2 to 12 

hours, relocate preventive maintenance 

surveillance requirements (SRs) to 

licensee controlled programs, provide 

alternate testing criteria for battery 

charger testing, replace battery specific 

gravity monitoring with float current 

monitoring, relocate and create a 

Section 5.5 program to reference actions 

for cell voltage and electrolyte level, and 

provide specific actions and increased 

completion times for out-of-limits 

conditions for certain battery 

parameters. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

This is a proposed change to the DC system 

Technical Specifications. No physical 

changes are being proposed to any system 

designed to prevent a previously evaluated 

accident, such as a loss of coolant accident 

with a loss of offsite power. 

This proposed TS change provides specific 

completion times for certain inoperable DC 

components, and relocates some surveillance 

requirements to owner controlled programs. 

Additionally, monitoring of specific gravity 

will be replaced with float current 

monitoring, and some Action levels for cell 

voltage and electrolyte level are relocated to 

owner controlled programs. 

The completion time for battery charger 

inoperability is increased to 7 days; however, 

only after verification that the associated 

battery is fully operable, without such 

verification, the 7 day completion time is not 

used. Thus, adequate DC to support design 

basis events is ensured. 

Increasing the station service battery out of 

service time from 2 to 12 hours will allow 

more time for proper maintenance to repair 

a faulty battery. However, the 12 hour out of 

service time is still a very restrictive time and 

so the probability of an event where the 

battery would be needed within this 12 hour 

time frame is very low. In fact, a probability 

risk assessment of the increased out of 

service time has been performed and it fell 

within the criteria of Reg Guide 1.174 and 

1.177. 

The relocation of certain SRs and action 

levels is done for surveillances and 

parameter action levels that are more 

intended to monitor and maintain long term 

Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 60687

component performance. These relocated 

items are not meant as clear levels at which 

the DC components can no longer be 

considered operable. Those that are remain 

in the TS. Additionally, this particular owner 

controlled program will be referenced in 

proposed Section 5.5.13 of the TS. This 

commitment to the program will insure that 

the DC system will continue to be adequately 

monitored and maintained. 

Therefore, this proposed change to the TS 

ensures that the DC system will be able to 

provide its safety function. The probability 

and consequences of a previously evaluated 

accident are thus not increased. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

None of the DC components will be 

physically altered. Furthermore, the design 

bases for the DC distribution systems, 

batteries and chargers is not changing. 

Although some surveillance requirements are 

being relocated and one (specific gravity 

monitoring) is being eliminated, DC system 

components will still be adequately 

surveilled and maintained. Therefore, no, 

new modes of operation or failure are 

introduced by the proposed TS change and 

therefore, the possibility of a new type event 

is not created. 

3. The proposed change does not involve 

a significant reduction in the margin of 

safety. 

The design functions of the DC system are 

unchanged. The proposed TS changes 

relocate many surveillance requirements and 

action levels to owner controlled programs. 

However, the owner controlled program is 

referenced in the new proposed Section 

5.5.13 of the TS. The SNC commitment to 

this program will continue to ensure that the 

DC system is adequately monitored, 

surveilled, and maintained to insure that it 

can perform its safety function when called 

upon. 

The addition of a 7 day completion time 

for the battery chargers can be used only if 

adequate battery capacity is verified. Thus, 

the DC system is capable of performing its 

safety function throughout the 7 day 

completion time. 

Increasing the allowed out of service time 

for the station service batteries does not 

result in a significant reduction in the margin 

of safety since the proposed 12 hour time 

limit is still a very short time. Probabilistic 

risk analysis shows that the core damage 

frequency and large early release fractions are 

within the guidelines of Reg Guides 1.174 

and 1.177. 

Elimination of specific gravity surveillance 

is acceptable since the float current 

monitoring adequately replaces it. 

For the above reasons, the proposed change 

does not involve a significant reduction in 

the margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ernest L. Blake, 

Jr., Esquire, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037. 

NRC Section Chief: Mary Jane RossLee, Acting. 

Tennessee Valley Authority, Docket 

Nos. 50–259, 50–260 and 50–296, 

Browns Ferry Nuclear Plant, Units 1, 2 

and 3, Limestone County, Alabama 

Date of amendment request: July 2, 

2004 (TS–449). 

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 July 2, 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 Technical 

Specification 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 

significance hazards consideration.

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Section Chief: Michael L. 

Marshall (Acting). 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant, Units 1 and 2, Hamilton 

County, Tennessee 

Date of amendment request: August 

12, 2004. 

Description of amendment request:

The proposed Technical Specification 

change will revise Surveillance 

Requirement 4.7.8.d.3 by removing the 

vacuum relief flow portion. The 

proposed revision removes criteria from 

the surveillance that is not necessary to 

verify the operability of the Auxiliary 

Building Gas Treatment System 

(ABGTS). The bases associated with the 

ABGTS will be revised to remove 

discussions regarding the vacuum relief 

flow portion of this surveillance as part 

of this effort. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed change removes an 

overly restrictive criterion for vacuum relief 

flow as part of the ABGTS operability 

verification. This criterion is not required for 

the verification of ABGTS operability and 

therefore, the removal does not reduce the 

associated safety function. No system 

modification or operating practices are 

changed by the proposed revision. The 

accident mitigation functions of the ABGTS 

will not be adversely affected by the 

proposed removal and offsite dose potential 

is not increased. 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? 

60688 Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 

No. The proposed change does not result 

in the alteration of plant equipment or 

components or the modification of operating 

requirements for plant systems. Additionally, 

the ABGTS functions serve to mitigate 

accident conditions and are not considered a 

source for accident generation. Therefore, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The proposed removal of an 

unnecessary criterion from the ABGTS 

surveillance will not result in a change to 

plant setpoints that function to maintain the 

safety margins. The ABGTS will continue to 

provide the required negative pressure 

conditions for the auxiliary building during 

accident conditions to maintain acceptable 

dose conditions. The actuation of safety 

features for accident mitigation will not be 

affected by the proposed changes. Therefore, 

the proposed change does not involve a 

significant reduction in a margin of safety.

The United States Nuclear Regulatory 

Commission (NRC) staff has reviewed 

the licensee’s analysis and, based on 

this review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Acting Section Chief: Michael 

Marshall, Jr. 

Tennessee Valley Authority (TVA), 

Docket Nos. 50–327 and 50–328, 

Sequoyah Nuclear Plant, Units 1 and 2, 

Hamilton County, Tennessee 

Date of amendment request: August 

18, 2004. 

Description of amendment request:

The proposed amendment would 

rename the Trip Setpoint column of 

Technical Specification (TS) Tables 2.2–

1 and 3.3–4, remove the inequality signs 

for the trip setpoint values as 

appropriate, and revise the inequality 

representation for the allowable values, 

as needed. This proposed amendment is 

a revision to a previous amendment 

request dated November 15, 2002 

(ADAMS Accession No. ML023290477), 

that supersedes the original request in 

its entirety. 

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:

A. The proposed amendment does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The proposed revisions for the nominal 

trip setpoint representation are 

administrative changes that will not impact 

the application of the reactor trip or ESF 

[engineered safety feature] actuation system 

instrumentation requirements. This is based 

on the setpoint requirements being applied 

without change, as well as the Avs [allowable 

values], in accordance with the current 

setpoint methodology. The removal of the 

inequalities associated with the trip setpoint 

values will be more appropriate for the use 

of nominal setpoint values but will not differ 

in application from the setpoint methodology 

utilized by TVA. Deletion of the nominal 

terminology associated with overtemperature 

delta temperature average temperature at 

rated thermal power (T′) provides a better 

representation of the limit associated with 

this value. In addition, this change will not 

alter plant equipment or operating practices. 

Therefore, the implementation of these 

changes will not increase the probability or 

consequences of an accident. 

The revision of the reactor coolant pump 

(RCP) underfrequency, intermediate range 

neutron flux P–6, and fuel storage pool area 

radiation monitor trip setpoints and the Avs 

for the RCP underfrequency, intermediate 

range neutron flux P–6, and undervoltage has 

been evaluated and the results are 

documented in approved calculations. These 

calculations verify that the revised values are 

acceptable in accordance with appropriate 

calculation methodologies and that they will 

continue to support the accident analysis. 

These revisions will not require changes to 

the instrumentation settings currently being 

used or the methods for maintaining them. 

The offsite dose potential will be reduced 

because the proposed TS values are more 

conservative and will ensure the adequacy of 

designed safety functions to limit the release 

of radioactivity. Therefore, the proposed 

revision of these values will not significantly 

increase the probability or consequences of 

an accident. 

B. The proposed amendment does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

The revision of the nominal trip setpoint 

representation and elimination of the 

nominal nomenclature, as well as the revised 

setpoint values and Avs will not alter the 

plant configuration or functions. The revised 

setpoints and the proposed operability limits 

will continue to provide acceptable initiation 

of safety functions for the mitigation of 

postulated accidents as required by the 

design basis. The primary function of the 

reactor protection system, the ESF actuation 

system, and the radiation monitoring 

function is to initiate accident mitigation 

functions. These functions are not considered 

to be initiators of postulated accidents. The 

proposed changes do not create the 

possibility of a new or different kind of 

accident because the design functions are not 

altered and the proposed values meet the 

accident analysis requirements for accident 

mitigation. 

C. The proposed amendment does not 

involve a significant reduction in a margin of 

safety. 

The setpoint and Av revisions proposed in 

this request were evaluated and found to be 

acceptable without impact to the safety limits 

required for the associated functions. The 

nominal trip setpoint representation change 

and the elimination of inappropriate nominal 

indications do not alter the TS functions or 

their application and will not require 

changes to design settings. Plant systems will 

continue to be actuated for those plant 

conditions that require the initiation of 

accident mitigation functions. The margin of 

safety is not reduced because the proposed 

conservative changes to the Av and setpoint 

representations will not change design 

functions and the initiation of accident 

mitigation functions for appropriate plant 

conditions is ensured.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Section Chief: Michael L. 

Marshall, Jr. (Acting). 

Notice of Issuance of Amendments to 

Facility Operating Licenses

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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 

Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 60689

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

Systems (ADAMS) Public Electronic 

Reading Room on the internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the NRC 

Public Document Room (PDR) Reference 

staff at 1–800–397–4209, 301–415–4737 

or by email to pdr@nrc.gov. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units Nos. 1, 2, and 

3, Maricopa County, Arizona 

Date of application for amendments:

August 22, 2003. 

Brief description of amendments: The 

amendments authorize revision of the 

Updated Final Safety Analysis Report to 

incorporate the description of the 

approved change to the maximum fuel 

pin pressurization criteria used in the 

evaluation of the design basis fuelhandling accident as described in the 

amendment application of August 22, 

2003. 

Date of issuance: September 27, 2004. 

Effective date: September 27, 2004, 

and shall be implemented within 60 

days of the date of issuance. 

Amendment Nos.: Unit 1–153, Unit 

2–153, Unit 3–153. 

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The 

amendments authorize the revision of 

the Updated Final Safety Analysis 

Report. 

Date of initial notice in Federal 

Register: December 9, 2003 (68 FR 

68656). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

September 27, 2004. 

No significant hazards consideration 

comments received: No. 

Carolina Power & Light Company, 

Docket No. 50–261, H.B. Robinson 

Steam Electric Plant, Unit No. 2, 

Darlington County, South Carolina 

Date of application for amendment:

May 10, 2002, as supplemented March 

12, 2003, April 10, 2003, March 5, 2004, 

and July 22, 2004. 

Brief description of amendment: The 

amendment approves full 

implementation of the alternative source 

term, with the exception of the loss-ofcoolant accident. 

Date of issuance: September 24, 2004. 

Effective date: September 24, 2004. 

Amendment No.: 201. 

Renewed Facility Operating License 

No. DPR–23: Amendment revises the 

Technical Specifications. 

Date of initial notice in Federal 

Register: April 1, 2003 (68 FR 15758). 

The April 10, 2003, March 5, 2004, and 

July 22, 2004, supplements 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 September 24, 

2004. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc., 

Docket No. 50–336, Millstone Power 

Station, Unit No. 2, New London 

County, Connecticut 

Date of application for amendment:

September 26, 2002, as supplemented 

June 2, 2003, May 7, June 18, and 

August 24, 2004. 

Brief description of amendment: The 

amendment revised the technical 

specifications (TSs) to allow relaxation 

of containment operability requirements 

while handling irradiated fuel and core 

alterations. 

Date of issuance: September 20, 2004.

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 284. 

Facility Operating License No. DPR–

65: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: November 12, 2002 (67 FR 

68731). The supplements dated June 2, 

2003, May 7, June 18, and August 24, 

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 September 20, 

2004. 

No significant hazards consideration 

comments received: No. 

Duke Energy Corporation, et al., Docket 

Nos. 50–413 and 50–414, Catawba 

Nuclear Station, Units 1 and 2, York 

County, South Carolina 

Date of application for amendments:

May 9, 2002, as supplemented by letter 

dated June 11, 2003, and August 18 and 

September 22, 2004. 

Brief description of amendments: The 

amendments approve changes to the 

Updated Final Safety Analysis Report 

for Catawba, Units 1 and 2 to eliminate 

the single failure of either of the 125 

VDC Distribution Centers, EDE or EDF, 

from the design-basis steam generator 

tube rupture accident analyses. 

Date of issuance: September 24, 2004. 

Effective date: As of the date of 

issuance and shall be implemented with 

the next update of the Safety Analysis 

Report in accordance with 10 CFR 

50.71(e) 

Amendment Nos.: 217, 211. 

Renewed Facility Operating License 

Nos. NPF–35 and NPF–52: Amendments 

revise the Licensing Basis. 

Date of initial notice in Federal 

Register: July 23, 2002 (67 FR 48215). 

The supplements dated June 11, 2003, 

and August 18 and September 22, 2004, 

provided clarifying information that did 

not change the scope of the May 9, 2002, 

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 24, 

2004. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2, 

Pope County, Arkansas 

Date of application for amendment:

June 30, 2003, as supplemented by 

letters dated November 20, 2003, 

February 27, 2004, and September 10, 

2004. 

Brief description of amendment: The 

amendment (1) reorganizes the Arkansas 

Nuclear One, Unit No. 2 (ANO–2) 

Technical Specification (TS) Section 

6.0, Administrative Controls, (2) 

modifies the ANO–2 Facility Operating 

License, and Actions and Surveillance 

Requirements (SRs) of various other 

TSs, to support the reorganization of 

Section 6.0, and (3) modifies several 

Actions and SRs that are related to 

systems that are shared by ANO–2 and 

Arkansas Nuclear One, Unit No. 1. 

Date of issuance: September 28, 2004. 

Effective date: As of the date of 

issuance to be implemented within 120 

days from the date of issuance. 

60690 Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 

Amendment No.: 255. 

Facility Operating License No. NPF–6:

Amendment revises the Facility 

Operating License and Technical 

Specifications.

Date of initial notice in Federal 

Register: December 9, 2003 (68 FR 

68663). The supplements dated 

February 27, 2004, and September 10, 

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 September 28, 

2004. 

No significant hazards consideration 

comments received: No. 

Florida Power and Light Company, 

Docket No. 50–335, St. Lucie Plant, Unit 

No. 1, St. Lucie County, Florida 

Date of application for amendment:

November 25, 2002, as supplemented by 

letters dated May 14, 2003, September 

29, 2003, and March 25, 2004. 

Brief description of amendment: The 

proposed amendment would revise 

Technical Specification 3.9.11, ‘‘Storage 

Pool Water Level’’ and TS 5.6.1, ‘‘Fuel 

Storage—Criticality.’’ This amendment 

permits St. Lucie Unit 1 to credit 

soluble boron, fuel loading restrictions, 

and control element assemblies in the 

spent fuel pool criticality analyses and 

eliminate the need to credit Boraflex 

neutron absorbing material for reactivity 

control. 

Date of Issuance: September 23, 2004. 

Effective Date: As of the date of 

issuance and shall be implemented by 

September 30, 2005. 

Amendment No.: 193. 

Renewed Facility Operating License 

No. DPR–67: Amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: January 7, 2003 (68 FR 806). 

The May 14, 2003, September 29, 2003, 

and March 25, 2004, supplements did 

not affect the original proposed no 

significant hazards determination, or 

expand the scope of the request as 

noticed in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 23, 

2004. 

No significant hazards consideration 

comments received: No. 

FPL Energy Seabrook, LLC, Docket No. 

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

Rockingham County, New Hampshire 

Date of amendment request: August 

25, 2003, as supplemented by letters 

dated February 9, February 23, March 

25, April 15, May 20, and July 29, 2004. 

Description of amendment request:

The amendment revised the Technical 

Specifications (TSs) to extend the 

emergency diesel generator allowed 

outage time from 72 hours to a period 

of 14 days, and to allow extension of the 

current two-hour time requirement to 

four hours for verification of redundant 

component operability. These changes 

are in support of installing a non-safetyrelated supplemental emergency power 

system. The Bases of the affected TSs 

will be modified to address the changes. 

Date of issuance: September 21, 2004. 

Effective date: As of its date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 97. 

Facility Operating License No. NPF–

86: The amendment revised the TSS. 

Date of initial notice in Federal 

Register: December 29, 2003 (68 FR 

68669). The supplements dated 

February 9, February 23, March 25, 

April 15, May 20, and July 29, 2004, 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 September 21, 

2004. 

No significant hazards consideration 

comments received: No. 

Indiana Michigan Power Company, 

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

C. Cook Nuclear Plant, Units 1 and 2, 

Berrien County, Michigan 

Date of application for amendments:

February 14, 2004, as supplemented 

July 26, 2004. 

Brief description of amendments: The 

amendments modify technical 

specification (TS) 3.9.2 limiting 

condition for operation, delete TS 

surveillance requirements (SRs) 4.9.2.a 

and 4.9.2.b for the Source Range 

Neutron Flux Monitor channel 

functional test, revise SR 4.9.2.c for the 

channel check test, and add a 

requirement to perform a channel 

calibration every 18 months as well as 

revise TS 4.10.4.2 and 4.10.3.2 (Units 1 

and 2 respectively) for Intermediate and 

Power Range channel functional test. 

Date of issuance: September 23, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 283, 267. 

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: May 11, 2004 (69 FR 26191). 

The supplemental letter contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination and did not 

expand the scope of the original Federal 

Register notice. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 23, 

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:

May 25, 2004, as supplemented August 

6, 2004. 

Brief description of amendment: The 

amendment revises technical 

specification (TS) 3.10.f.2 to add an 

allowed outage time for the individual 

rod position indication (IRPI) system of 

24 hours with more than one IRPI group 

inoperable and adds the definition of 

‘‘immediately’’ to TS Section 1.0. 

Date of issuance: September 22, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 176. 

Facility Operating License No. DPR–

43: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 6, 2004 (69 FR 40675). 

The supplement dated August 6, 

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 Safety Evaluation dated 

September 22, 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 7, 2003, as supplemented March 17, 

May 18, and August 18, 2004.

Brief description of amendment: The 

amendment adds Technical 

Specification Section 3.3.e.1.A.3, which 

provides requirements for turbine 

building service water header isolation 

logic. 

Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 60691

Date of issuance: September 24, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 177. 

Facility Operating License No. DPR–

43: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 5, 2003 (68 FR 46244). 

The supplements dated March 17, 

May 18, and August 18, 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 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:

July 24, 2003. 

Brief description of amendments: The 

amendments revise Technical 

Specification (TS) 3.8.4, ‘‘DC Sources—

Operating,’’ TS 3.8.5, ‘‘DC Sources—

Shutdown,’’ and TS 3.8.6, ‘‘Battery Cell 

Parameters,’’ and add a new TS 5.5.17, 

‘‘Battery Monitoring and Maintenance 

Program.’’ The changes adopt in part the 

NRC-approved Technical Specification 

Task Force (TSTF–360, Revision 1, ‘‘DC 

Electrical Rewrite.’’

Date of issuance: September 20, 2004. 

Effective date: September 20, 2004, 

and shall be implemented within 120 

days from the date of issuance. The 

licensee shall reflect the relocation of 

TS requirements to licensee-controlled 

programs and the TS Bases, as described 

in the licensee’s letter dated July 24, 

2003, and the NRC safety evaluation 

attached to the amendment, in the next 

scheduled update of the Final Safety 

Analysis Report Update submitted 

pursuant to 10 CFR 50.71(e). 

Amendment Nos.: Unit 1—172; Unit 

2—174. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: September 2, 2003 (68 FR 

52236). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 20, 

2004. 

No significant hazards consideration 

comments received: No. 

PPL Susquehanna, LLC, Docket No. 50–

388, Susquehanna Steam Electric 

Station, Unit 2 (SSES–2), Luzerne 

County, Pennsylvania 

Date of application for amendments:

September 16, 2003, as supplemented 

by letter dated April 27, 2004. 

Brief description of amendments: The 

amendment revised the values of the 

Safety Limit for Minimum Critical 

Power Ratio in TS 2.1.1.2 for current 

SSES–2 Cycle 12 mid-cycle tworecirculation-loop and singlerecirculation-loop operation. 

Date of issuance: September 21, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 191. 

Facility Operating License No. NPF–

22: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: October 28, 2003 (68 FR 

61480). The supplement dated April 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 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 21, 2004. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear, LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of application for amendments:

July 29, 2002, as supplemented by 

letters dated March 28, 2003, May 1, 

2003, and August 20, 2004. 

Brief description of amendments: The 

amendments modify the Salem Nuclear 

Generating Station, Unit Nos. 1 and 2, 

Technical Specifications (TSs) 

requirements for containment closure 

associated with the equipment hatch 

and personnel airlocks during Core 

Alterations and movement of irradiated 

fuel within the containment. The 

change allows the equipment hatch and 

the personnel airlocks to remain open 

during fuel movement inside 

containment provided administrative 

controls are in place to ensure the 

closure of the equipment hatch and 

personnel airlock following a fuel 

handling accident within the 

containment building. In addition, the 

associated TS Bases are revised. 

Date of issuance: September 16, 2004. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 90 days. 

Amendment Nos.: 263 and 245. 

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: August 20, 2002 (67 FR 

53989). The licensee’s supplements 

dated March 28, 2003, May 1, 2003, and 

August 20, 2004, provided clarifying 

information that did not change the 

scope of the proposed amendments as 

described in the original notice of 

proposed action published in the 

Federal Register, and did not change 

the initial proposed no significant 

hazards consideration determination. 

The Commission’s related evaluation of 

the amendments is contained in a Safety 

Evaluation dated September 16, 2004. 

No significant hazards consideration 

comments received: No. 

Rochester Gas and Electric Corporation, 

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

Power Plant, Wayne County, New York 

Date of application for amendment:

April 9, 2002, as supplemented January 

10, 2003, February 24, 2004, and August 

27, 2004. 

Brief description of amendment: The 

amendment revised the Ginna Improved 

Technical Specification with regards to: 

relocating figures associated with Core 

Safety Limits to the Core Operating 

Limits Report (COLR), relocating 

Overtemperture DT and Overpower DT 

parameters to the COLR, and replacing 

current trip setpoints for the Reactor 

Protection System and the Engineered 

Safety Feature Actuation System with 

Limiting Safety System Settings in 

accordance with the Instrument Society 

of America Standard 67.04, Part 2. 

Date of issuance: September 22, 2004. 

Effective date: As of the date of 

issuance to be implemented within 1 

year. 

Amendment No.: 85. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: May 28, 2002 (67 FR 36933). 

The supplements dated January 10, 

2003, February 24, 2004 and August 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 staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 22, 

2004. 

No significant hazards consideration 

comments received: No.

60692 Federal Register / Vol. 69, No. 196 / Tuesday, October 12, 2004 / Notices 

1 15 U.S.C. 78s(b)(1).

2 17 CFR 240.19b–4.

Southern California Edison Company, et 

al., Docket No. 50–206, San Onofre 

Nuclear Generating Station, Unit 1, San 

Diego County, California 

Date of application for amendment:

January 28, 2004, supplemented by a 

letter dated July 23, 2004. 

Brief description of amendment: The 

amendment revises the SONGS Unit 1 

License and Permanently Defueled 

Technical Specifications to modify or 

remove operational and administrative 

requirements that are not applicable 

upon the transfer of all spent fuel from 

the spent fuel pool into the SONGS dry 

cask storage Independent Spent Fuel 

Storage Installation. 

Date of issuance: September 21, 2004. 

Effective date: As of the date that all 

reactor fuel has been permanently 

removed from the spent fuel pool and 

stored in an Independent Spent Fuel 

Storage Installation. The license 

amendment shall be implemented 

within 30 days of its effective date. 

Amendment No.: 163. 

Facility Operating License No. DPR–

13: This amendment revises both the 

license and the technical specifications. 

Date of initial notice in Federal 

Register: March 30, 2004 (69 FR 

16623). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

September 21, 2004. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket 

Nos. 50–259, 50–260, and 50–296, 

Browns Ferry Nuclear Plant, Units 1, 2, 

and 3, Limestone County, Alabama 

Date of application for amendments:

July 31, 2002 as supplemented by letters 

dated December 9, 2002, February 12, 

2003, March 26, 2003, July 11, 2003, 

July 17, 2003, May 17, 2004, July 2, 

2004, August 24, 2004 and September 

17, 2004. 

Description of amendment request:

The amendments requested full 

implementation of an alternative source 

term (AST) methodology for the Units 1, 

2, and 3 operating licenses and design 

bases. The amendments adopt the AST 

methodology by revising the current 

accident source term and replacing it 

with an accident source term as 

prescribed in 10 CFR 50.67. The 

submittals also proposed to revise and/

or remove the Technical Specification 

(TS) Sections associated with control 

room emergency ventilation (CREV), 

standby gas treatment (SGT), standby 

liquid control (SLC), and secondary 

containment systems. Additionally, the 

submittals requested modification of the 

licensing and design basis to reflect the 

application of the AST methodology 

and the function of the SLC system, and 

deletion of a license condition for Units 

2 and 3. 

The supplements to the original 

application included the withdrawal of 

the request to delete one of the TS 

Sections described above, associated 

with the absorption of elemental iodine 

by the SGT and CREV systems charcoal 

filters. Also the supplements added a 

new TS Section to require verification 

that the minimum fuel decay period has 

passed prior to moving fuel after the 

reactor is shut down. 

Date of issuance: September 27, 2004. 

Effective date: Date of issuance, to be 

implemented prior to restart of Unit 1, 

and within 120 days for Units 2 and 3. 

Amendment Nos.: 251, 290 and 249. 

Facility Operating License Nos. DPR–

33, DPR–52, and DPR–68: Amendments 

revised the Operating Licenses and TSs. 

Date of initial notice in Federal 

Register: October 15, 2002 (67 FR 

63697). The supplements dated 

December 9, 2002, February 12, March 

26, July 11, and July 17, 2003, provided 

information that changed the scope of 

the original request, therefore another 

Federal Register notice was published 

on April 27, 2004 (69 FR 22883). 

However, the supplements dated May 

17, July 2, August 24, and September 17, 

2004, provided clarifying information 

that did not expand the scope of the 

revised request or the proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 27, 

2004. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant, Units 1 and 2, Hamilton 

County, Tennessee 

Date of application for amendment:

March 5, 2004. 

Brief description of amendment: The 

amendments delete surveillance 

requirements to perform certain channel 

functional tests of the source range, 

intermediate, and power range neutron 

flux monitors. These amendments 

eliminate extraneous and unnecessary 

performance of these surveillances. 

Date of issuance: September 20, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days. 

Amendment Nos.: 295 and 285. 

Facility Operating License No. DPR–

77 and DPR–79: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: April 13, 2004 (69 FR 19576). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated September 20, 2004. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant, Units 1 and 2, Hamilton 

County, Tennessee 

Date of application for amendment:

March 5, 2004. 

Brief description of amendment: The 

amendments eliminate the requirements 

in the technical specifications 

associated with hydrogen recombiners 

and hydrogen monitors. 

Date of issuance: September 20, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days. 

Amendment Nos.: 296 and 286. 

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revised 

the technical specifications. 

Date of initial notice in Federal 

Register: April 13, 2004 (69 FR 19576). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 20, 

2004. 

No significant hazards consideration 

comments received: No.

Dated at Rockville, Maryland, this 1st day 

of October, 2004.

For the Nuclear Regulatory Commission. 

William H. Ruland, 

Acting Director, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation.

[FR Doc. 04–22544 Filed 10–8–04; 8:45 am] 

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE 

COMMISSION 

[Release No. 34–50478; File No. SR–NASD–

2004–107] 

Self-Regulatory Organizations; 

National Association of Securities 

Dealers, Inc.; Notice of Filing of a 

Proposed Rule Change Relating to 

Computer Generated Quoting in 

Exchange-Listed Securities 

September 30, 2004. 

Pursuant to Section 19(b)(1) of the 

Securities Exchange Act of 1934 

(‘‘Act’’) 1 and Rule 19b–4 thereunder,2

notice is hereby given that on July 12, 

2004, the National Association of 

Securities Dealers, Inc. (‘‘NASD’’), 

through its subsidiary, The Nasdaq 

Stock Market, Inc. (‘‘Nasdaq’’), filed