57978 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

Office of Nuclear Material Safety and 

Safeguards, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001.

CONDUCT OF THE MEETING: Dr. Cerqueira, 

M.D., will chair the meeting. Dr. 

Cerqueira will conduct the meeting in a 

manner that will facilitate the orderly 

conduct of business. The following 

procedures apply to public participation 

in the meeting: 

(1) Persons who wish to provide a 

written statement should submit a 

reproducible copy to Angela McIntosh, 

U.S. Nuclear Regulatory Commission, 

Two White Flint North, Mail Stop T8F5, 

Washington, DC 20555–0001. Hard copy 

submittals must be postmarked by 

September 29, 2004. Electronic 

submittals must be submitted by 

October 1, 2004. Any submittal must 

pertain to the topic on the agenda for 

the meeting. 

(2) Questions from members of the 

public will be permitted during the 

meeting, at the discretion of the 

Chairman. 

(3) The transcript and written 

comments will be available for 

inspection on NRC’s Web site (http://

www.nrc.gov) and at the NRC Public 

Document Room, 11555 Rockville Pike, 

Rockville, MD 20852–2738, telephone 

(800) 397–4209, on or about November 

12, 2004. Minutes of the meeting will be 

available on or about December 17, 

2004. 

This meeting will be held in 

accordance with the Atomic Energy Act 

of 1954, as amended (primarily Section 

161a); the Federal Advisory Committee 

Act (5 U.S.C. App); and the 

Commission’s regulations in Title 10, 

U.S. Code of Federal Regulations, Part 7.

Dated at Rockville, Maryland, this 21st day 

of September, 2004. 

For the Nuclear Regulatory Commission. 

Andrew L. Bates, 

Advisory Committee Management Officer.

[FR Doc. 04–21653 Filed 9–27–04; 8:45 am] 

BILLING CODE 7590–01–P

NUCLEAR REGULATORY 

COMMISSION 

Meetings; Sunshine Act

DATE: Weeks of September 27, October 

4, 11, 18, 25, November 1, 2004.

PLACE: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland.

STATUS: Public and closed.

MATTERS TO BE CONSIDERED: 

Week of September 27, 2004

There are no meetings scheduled for 

the week of September 27, 2004. 

Week of October 4, 2004—Tentative 

Thursday, October 7, 2004

9:25 a.m. Affirmation Session 

(Public Meeting) (Tentative). 

a. State of Alaska Department of 

Transportation and Public Facilities 

(Confirmatory Order Modifying 

License); appeals of LBP–04–16 by NRC 

Staff and Licensee (Tentative). 

b. Private Fuel Storage (Independent 

Spent Fuel Storage Installation) Docket 

No. 72–22–ISFSI (Tentative). 

c. USEC, Inc. (Tentative). 

10:30 a.m. Discussion of Security 

Issues (Closed—Ex. 1). 

1 p.m. Discussion of Security Issues 

(Closed—Ex. 1). 

2:30 p.m. Discussion of Security 

Issues (Closed—Ex. 1). 

Week of October 11, 2004—Tentative 

Wednesday, October 13, 2004

9:30 a.m. Briefing on 

Decommissioning Activities and Status 

(Public Meeting) (Contact: Claudia 

Craig, 301–415–7276). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov. 

1:30 p.m. Discussion of 

Intragovernmental Issues (Closed—Ex. 1 

& 9). 

Week of October 18, 2004—Tentative 

There are no meetings scheduled for 

the week of October 18, 2004. 

Week of October 25, 2004—Tentative 

There are no meetings scheduled for 

the week of October 25, 2004. 

Week of November 1, 2004—Tentative 

There are no meetings scheduled for 

the week of November 1, 2004. 

* The schedule for Commission 

meetings is subject to change on short 

notice. To verify the status of meetings 

call (recording)—(301) 415–1292. 

Contact person for more information: 

Dave Gamberoni, (301) 415–1651.

* * * * *

The NRC Commission meeting 

Schedule can be found on the Internet 

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

policy-making/schedule.html

* * * * *

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g.

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

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

301–4152100, or by e-mail at 

aks@nrc.gov. Determinations on 

requests for reasonable accommodation 

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

* * * * *

This notice is distributed by mail to 

several hundred subscribers; if you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301–415–1969). 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to dkw@nrc.gov.

Dated: September 23, 2004. 

Dave Gamberoni, 

Office of the Secretary.

[FR Doc. 04–21767 Filed 9–24–04; 9:34 am] 

BILLING CODE 7590–01–M

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments To Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a.(2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued, from September 

3, 2004, through September 16, 2004. 

The last biweekly notice was published 

on September 14, 2004 (69 FR 55466). 

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 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57979

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 O1F21, 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 basis 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/

requestor to relief. A petitioner/

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

A request for a hearing or a petition 

for leave to intervene must be filed by: 

(1) First class mail addressed to the 

57980 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

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 O1F21, 11555 

Rockville Pike (First Floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the NRC 

PDR Reference staff at 1–800–397–4209, 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units 1, 2, and 3, 

Maricopa County, Arizona 

Date of amendments request: July 9, 

2004.

Description of amendments request:

The amendments would revise the 

Technical Specifications (TSs) to allow 

operation of Palo Verde Nuclear 

Generating Station (PVNGS), Units 1 

and 3 up to a maximum reactor core 

power level of 3990 Megawatts thermal 

(MWt), an increase of 2.94 percent 

above the current licensed power level 

of 3876 MWt. The proposed 

amendments would also make 

administrative changes to the PVNGS 

Unit 2 TSs so that the changed pages 

would apply to the three PVNGS units. 

Operation at the uprated power level 

with replacement steam generators has 

been approved for PVNGS Unit 2. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

(1) Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Response: No. 

(a) Evaluation of the Probability of 

Previously Evaluated Accidents 

Plant Structures, Systems and Components 

(SSCs) have been verified to be capable of 

performing their intended design functions at 

uprated power conditions. Where necessary, 

a small number of minor modifications will 

be made prior to implementation of uprated 

power operations so that surveillance test 

acceptance criteria continues to be met. The 

analysis has concluded that operation at 

uprated power conditions will not adversely 

affect the capability or reliability of plant 

equipment. Current technical specification 

(TS) surveillance requirements ensure 

frequent and adequate monitoring of system 

and component operability. All systems will 

continue to be operated within current 

operating requirements at uprated 

conditions. Therefore, no new structure, 

system or component interactions have been 

identified that could lead to an increase in 

the probability of any accident previously 

evaluated in the Updated Final Safety 

Analysis Report (UFSAR). 

(b) Evaluation of the Consequences of 

Previously Evaluated Accidents 

The radiological consequences were 

reviewed for all design basis accidents 

(DBAs) (i.e., both LOCA [loss-of-coolant 

accident] and non-LOCA accidents) 

previously analyzed in the UFSAR. The 

analysis showed that the resultant 

radiological consequences for both LOCA 

and non-LOCA accidents remain either 

unchanged or have not significantly 

increased due to operation at uprated power 

conditions. The radiological consequences of 

all DBAs continue to meet established 

regulatory limits. 

(2) Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Response: No. 

The configuration, operation and accident 

response of the PVNGS [Palo Verde Nuclear 

Generating Station] Units I and 3 structures, 

systems, and components are unchanged by 

operation at uprated power conditions or by 

the associated proposed TS changes. 

Analyses of transient events have confirmed 

that no transient event results in a new 

sequence of events that could lead to a new 

accident or different scenario. 

The effect of operation at uprated power 

conditions on plant equipment has been 

evaluated. No new operating mode, safetyrelated equipment lineup, accident scenario, 

or equipment failure mode was identified as 

a result of operating at uprated conditions. In 

addition, operation at uprated power 

conditions does not create any new failure 

modes that could lead to a different kind of 

accident. Minor plant modifications, to 

support Implementation of uprated power 

conditions, will be made as required to 

existing SSCs. The basic design function of 

all SSCs remains unchanged and no new 

equipment or systems have been installed 

that could potentially introduce new failure 

modes or accident sequences. 

Based on this analysis, it is concluded that 

no new accident scenarios, failure 

mechanisms or limiting single failures are 

introduced as a result of the proposed 

changes. The proposed changes do not have 

an adverse effect on any safety-related system 

or design basis function. Therefore, the 

proposed changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

(3) Does the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

A comprehensive analysis was performed 

to evaluate the effects of power uprate on 

PVNGS Units 1 and 3. This analysis 

identified and defined the major input 

parameters to the NSSS [nuclear steam 

supply system], reviewed NSSS design 

transients, and reviewed the capabilities of 

the NSSS and BOP [balance of plant] fluid 

systems, NSSS/BOP interfaces, NSSS and 

BOP control systems, and NSSS and BOP 

SSCs. All appropriate NSSS accident 

analyses were re-performed to confirm that 

acceptable results were maintained and that 

the radiological consequences remained 

within regulatory and Standard Review Plan 

(SRP) limits. The nuclear and thermal 

hydraulic performance of nuclear fuel was 

also reviewed to confirm acceptable results. 

The analyses confirmed that all NSSS and 

BOP SSCs are capable, some with minor 

modifications, to safely support operations at 

uprated power conditions. 

The margin of safety of the reactor coolant 

pressure boundary is maintained under 

uprated power conditions. The design 

pressure of the reactor pressure vessel and 

reactor coolant system will not be challenged 

as the pressure mitigating systems were 

confirmed to be sufficiently sized to 

adequately control pressure under uprated 

power conditions. 

Reanalysis of containment structural 

integrity under Design Basis Accident (DBA) 

conditions indicates that the calculated peak 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57981

containment pressure (Pa) increases from 

52.0 psig [pounds per square inch gauge] to 

58.0 psig, but remains less than the 

containment internal design pressure of 60 

psig. The proposed value for Pa has been 

rounded up from the actual calculated value 

of 57.85 psig. 

Radiological consequences of the following 

accidents were reviewed: Main Steam Line 

Break, Locked Reactor Coolant Pump (RCP) 

Rotor, CEA Ejection, Small Steam Line Break 

Outside Containment, Steam Generator Tube 

Rupture, LBLOCA, SBLOCA, Waste Gas 

Decay Tank Rupture, Liquid Waste Tank 

Failure, and Fuel Handling Accident. The 

resultant radiological consequences for each 

of these accidents did not show a significant 

change due to uprated power conditions and 

10 CFR 100 and SRP limits continue to be 

met. 

The analyses supporting operation at 

power uprate conditions have demonstrated 

that all systems and components are capable 

of safely operating at uprated power 

conditions. All design basis accident 

acceptance criteria will continue to be met. 

Therefore, it is concluded that the proposed 

changes do not involve a significant 

reduction in the margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on that 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the request 

for amendments involves no significant 

hazards consideration. 

Attorney for licensee: Kenneth C. 

Manne, Senior Attorney, Arizona Public 

Service Company, P.O. Box 52034, Mail 

Station 7636, Phoenix, Arizona 85072–

2034. 

NRC Section Chief: Robert Gramm.

Carolina Power & Light Company, 

Docket No. 50–261, H.B. Robinson 

Steam Electric Plant, Unit No. 2, 

Darlington County, South Carolina 

Date of amendment request: August 

19, 2004. 

Description of amendment request:

The proposed amendment would revise 

the reactor coolant system (RCS) 

pressure and temperature limits by 

replacing Technical Specification 

Section 3.4.3, ‘‘RCS Pressure and 

Temperature (P/T) Limits,’’ Figures 

3.4.3–1 and 3.4.3–2, with figures that 

are applicable up to 35 effective fullpower years (EFPY). 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

(1) Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

The proposed RCS P/T limits are based on 

NRC-approved methodology and will 

continue to maintain appropriate limits for 

the HBRSEP [H.B. Robinson Steam Electric 

Plant], Unit No. 2, RCS up to 35 EFPY. These 

changes provide appropriate limits for 

pressure and temperature during heatup and 

cooldown of the RCS, thus ensuring that the 

probability of RCS failure is maintained 

acceptably low. These limits are not directly 

related to the consequences of accidents. 

Therefore, 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 previously evaluated? 

The proposed changes will continue to 

ensure that the RCS will be maintained 

within appropriate pressure and temperature 

limits during heatup and cooldown. No 

physical changes to the HBRSEP, Unit No. 2, 

systems, structures, or components are being 

implemented. There are no new or different 

accident initiators or sequences being created 

by the proposed Technical Specifications 

changes. Therefore, these changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

(3) Do the proposed changes involve a 

significant reduction in the margin of safety? 

The proposed changes ensure that the 

margin of safety for the fission product 

barriers protected by these functions will 

continue to be maintained. This conclusion 

is based on use of the applicable NRCapproved methodology for developing and 

establishing the proposed RCS P/T limits. 

Therefore, these changes do not involve a 

significant reduction in the margin of safety. 

Based on the preceding discussion, the 

requested change does not involve a 

significant hazards consideration.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Steven R. Carr, 

Associate General Counsel—Legal 

Department, Progress Energy Service 

Company, LLC, Post Office Box 1551, 

Raleigh, North Carolina 27602. 

NRC Section Chief: Michael Marshall 

(Acting). 

Connecticut Yankee Atomic Power 

Company, Docket No. 50–213, Haddam 

Neck Plant, Middlesex County, 

Connecticut 

Date of amendment request: August 

11, 2004. 

Description of amendment requests:

The Haddam Neck Plant (HNP) is 

currently undergoing active 

decommissioning. The proposed 

amendment would revise Technical 

Specifications (TS) to reflect removal of 

all Spent Nuclear Fuel (SNF) from the 

HNP spent fuel pool, and delete the 

requirement for submittal of an annual 

Occupational Radiation Exposure 

Report consistent with Industry’s 

Technical Specifications Task Force 

(TSTF)–369, Revision 1. 

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:

In accordance with 10 CFR 50.92, CYAPCO 

has reviewed the proposed changes and 

concluded that the proposed changes do not 

involve a Significant Hazard Consideration 

(SHC). The following is provided in support 

of this conclusion: 

Incorporation of TSTF–369, Revision 1:

CYAPCO has reviewed the no significant 

hazards consideration determination 

published in the Federal Register (69 FR 

35067) as part of the CLIIP. CYAPCO has 

concluded that the determination presented 

in the Federal Register is applicable to the 

HNP and is hereby incorporated by reference 

to satisfy the requirements of 10 CFR 50.91. 

Deletion and Relocation of Technical 

Specifications: The proposed changes do not 

involve an SHC because the changes would 

not: 

(1) Involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The proposed changes (deletion of 

operational requirements and certain design 

requirements) reflect the complete transfer of 

the spent fuel from the spent fuel pool to the 

Independent Spent Fuel Storage Installation 

(ISFSI). Design basis accidents related to the 

spent fuel pool are discussed in the Haddam 

Neck Plant (HNP) Updated Final Safety 

Analysis (UFSAR) Chapter 15. These 

postulated accidents are predicated on spent 

fuel being stored in the spent fuel pool. With 

the removal of the spent fuel from the spent 

fuel pool, there are no remaining safety 

related Structures, Systems, and Components 

(SSCs) to be monitored and there are no 

credible accidents that require the actions of 

a Certified Fuel Handler or an Equipment 

Operator to prevent occurrence or mitigate 

the consequences of an accident. 

In addition, the HNP UFSAR Chapter 15 

also provides a discussion of other 

radiological events postulated to occur as a 

result of decommissioning with the bounding 

consequences resulting from a fire in a resin 

container. The proposed changes do not have 

an adverse impact on decommissioning 

activities or any of their postulated 

consequences. 

The proposed changes related to the 

relocation of certain administrative 

requirements do not affect operating 

procedures or administrative controls that 

have the function of preventing or mitigating 

any design basis accidents. In addition, these 

proposed changes are consistent with the 

guidance of NRC Administrative Letter 95–

06. 

Therefore, the proposed changes do not 

involve a significant increase in the 

57982 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

probability or consequences of any accident 

previously evaluated. 

(2) Create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

The proposed changes eliminate the 

operational requirements and certain design 

requirements associated with the storage of 

the spent fuel in the spent fuel pool, and 

relocate certain administrative controls to the 

Connecticut Yankee Quality Assurance 

Program (CYQAP). With the complete 

removal of the spent fuel from the spent fuel 

pool, there are no safety related SSCs that 

remain at the plant. Thus the proposed 

changes will not have any effect on the 

operation or design function of safety related 

SSCs. The proposed changes do not 

introduce any new failure modes. Therefore, 

the proposed changes will not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

(3) Involve a significant reduction in a 

margin of safety. 

The design basis and accident assumptions 

within the HNP UFSAR and the Technical 

Specifications relating to spent fuel are no 

longer applicable. The proposed changes do 

not affect remaining plant operations, 

systems, or components supporting 

decommissioning activities. In addition, the 

proposed changes do not result in a change 

in initial conditions, system response time, 

or in any other parameter affecting the course 

of a decommissioning activity accident 

analysis. Therefore, the proposed changes do 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the licensee’s 

analysis and, based on this review, it appears 

that the three standards of 10 CFR 50.92(c) 

are satisfied. Therefore, the NRC staff 

proposes to determine that the amendment 

request involves no significant hazards 

consideration.

NRC Section Chief: Claudia Craig. 

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 amendment request: May 27, 

2004. 

Description of amendment request:

The proposed amendment would delete 

the requirements from the Technical 

Specifications (TS) to maintain 

hydrogen recombiners and hydrogen 

monitors. A notice of availability for the 

TS improvement using the consolidated 

line item improvement process was 

published in the Federal Register on 

September 25, 2003 (68 FR 554416). 

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 LightWater-Cooled 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, ‘‘Standards 

for Combustible Gas Control System in 

Light-Water-Cooled Power Reactors,’’

eliminated the requirements for 

hydrogen recombiners 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 

May 27, 2004. Basis for proposed no 

significant hazards consideration 

determination: As required by 10 CFR 

50.91(a), an analysis of the issue of no 

significant hazards consideration is 

presented below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The 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 monitors 

are no longer required to mitigate designbasis accidents and, therefore, the hydrogen 

monitors do not meet the definition of a 

safety-related component as defined in 10 

CFR 50.2. Category 1 in RG 1.97 is intended 

for key variables that most directly indicate 

the accomplishment of a safety function for 

design-basis accident events. The hydrogen 

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.

The regulatory requirements for the 

hydrogen 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, and 

removal of the hydrogen monitors from TS 

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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TSs, will not result in any failure mode 

not previously analyzed. The hydrogen 

recombiner and hydrogen monitor equipment 

was intended to mitigate a design-basis 

hydrogen release. The hydrogen recombiner 

and hydrogen 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, 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 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57983

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. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

Removal of hydrogen monitoring from TSs 

will not result in a significant reduction in 

their functionality, reliability, and 

availability.

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Lisa F. 

Vaughn, Legal Department (PB05E), 

Duke Energy Corporation, 422 South 

Church Street, Charlotte, North Carolina 

28201–1006. 

NRC Section Chief: Mary Jane RossLee (Acting). 

Duke Energy Corporation, Docket Nos. 

50–269, 50–270, and 50–287, Oconee 

Nuclear Station, Units 1, 2, and 3, 

Oconee County, South Carolina 

Date of amendment request: August 

20, 2004. 

Description of amendment request:

The proposed amendments would 

revise Technical Specification (TS) 

3.3.8, ‘‘Post Accident Monitoring [PAM] 

Instrumentation,’’ to eliminate TS 

requirements associated with the reactor 

building spray (RBS) flow instruments 

commensurate with the importance of 

their revised post-accident function. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

Criterion 1—The proposed amendment 

would not involve a significant increase in 

the probability or consequences of an 

accident previously evaluated 

Duke proposes to remove the RBS flow 

instrument from Technical Specification 

Table 3.3.8–1 based on a change in its 

purpose due to recent modifications 

completed at Oconee. The TS 3.3.8 

requirement to declare the affect [affected] 

RBS System train inoperable is conservative 

(and inappropriate) when the associated RBS 

flow instrument is inoperable. Due to recent 

plant modifications, the RBS flow 

instruments are no longer needed to allow 

the operator to throttle flow to preclude RBS 

pump runout post accident. The revised post 

accident function of this PAM instrument is 

to provide information to indicate the 

operation of the RBS System. There are 

alternate means to verify that the RBS is in 

operation, such as, verifying the RBS pump 

and valve status. The failure of an RBS flow 

instrument has no impact on the probability 

of an accident analyzed in the UFSAR 

[Updated Final Safety Analysis Report]. The 

RBS flow instrument is no longer needed to 

mitigate the consequences of an accident 

analyzed in the UFSAR. As such, the 

proposed LAR [license amendment request] 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

Criterion 2—The proposed amendment 

would not create the possibility of a new or 

different kind of accident from any kind of 

accident previously evaluated 

Duke proposes to remove the RBS flow 

instrument from Technical Specification 

Table 3.3.8–1 based on a change in its 

purpose due to recent modifications 

completed at Oconee. The TS 3.3.8 

requirement to declare the affect [affected] 

RBS System train inoperable is conservative 

(and inappropriate) when the associated RBS 

flow instrument is inoperable. Due to recent 

plant modifications, the RBS flow 

instruments are no longer needed to allow 

the operator to throttle flow to preclude RBS 

pump runout post accident. These changes 

do not alter the nature of events postulated 

in the Safety Analysis Report nor do they 

introduce any unique precursor mechanisms. 

Therefore, the proposed amendment will not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

Criterion 3—The proposed amendment 

would not involve a significant reduction in 

a margin of safety 

The proposed TS changes do not 

unfavorably affect any plant safety limits, set 

points, or design parameters. The changes 

also do not unfavorably affect the fuel, fuel 

cladding, RCS [reactor coolant system], or 

containment integrity. Therefore, the 

proposed TS change, which changes TS 

requirements associated with revised PAM 

function of the RBS flow instrument 

channels, 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: Anne W. 

Cottingham, Winston and Strawn LPP, 

1400 L Street, NW., Washington, DC 

20005. 

NRC Section Chief: Mary Jane RossLee, Acting. 

Duke Energy Corporation, Docket Nos. 

50–269, 50–270, and 50–287, Oconee 

Nuclear Station, Units 1, 2, and 3, 

Oconee County, South Carolina 

Date of amendment request: August 

26, 2004. 

Description of amendment request:

The proposed amendments would add 

new Technical Specification (TS) 3.3.29 

and TS Bases 3.3.29, ‘‘Reactor Building 

Auxiliary Cooler (RBAC) Isolation 

Circuitry,’’ to accommodate new 

circuitry that isolates non-safety 

portions of the low pressure service 

water (LPSW) system piping inside 

containment that supply the RBACs. 

This isolation eliminates potentially 

damaging water hammers that could 

occur in the event of certain designbases events or transients. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

Criterion 1—The Proposed Amendment 

Would Not Involve a Significant 

Increase in the Probability or 

Consequences of an Accident Previously 

Evaluated 

The requested license amendment 

would add a new Technical 

Specification to provide appropriate 

controls for the Reactor Building (RB) 

Auxiliary Cooler (RBAC) isolation 

circuitry that is being added to the 

design of the three Oconee units. The 

RBAC isolation circuitry provides an 

automatic means to isolate the LPSW 

flow stream to the RBACs on a loss of 

LPSW flow that can lead to a column 

closure water hammer inside the RB 

when LPSW flow is restarted. The new 

circuitry ensures that significant 

waterhammers do not occur in the 

LPSW piping to the RBACs and other 

RB components. The new circuitry will 

eliminate an Operable but degraded/

non-conforming condition associated 

with potentially damaging 

waterhammers. 

The proposed RBAC isolation 

circuitry Technical Specification will 

provide means to assure that the RBAC 

isolation circuitry operates at a 

performance level necessary to provide 

for safe operation of the LPSW system 

following installation of the LPSW 

modification and RBAC isolation 

circuitry at each of the three units. The 

addition of the RBAC isolation circuitry 

Technical Specification does not 

increase the probability or consequences 

of any accident previously evaluated. 

Criterion 2—The Proposed Amendment 

Would Not Create the Possibility of a 

New or Different Kind of Accident From 

Any Accident Previously Evaluated 

The proposed RBAC isolation 

circuitry Technical Specification 

provides a means to assure the isolation 

circuitry operates at a performance level 

necessary to provide for safe operation 

57984 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

of the modified LPSW system flow to 

the RBACs. The change enhances the 

plant design by eliminating the 

possibility of significant waterhammers 

that could occur inside the RB on a loss 

of LPSW flow to the RBACs. 

The proposed Technical Specification 

will not create the possibility of a new 

or different kind of accident from any 

kind of accident previously evaluated. 

Criterion 3—The Proposed Amendment 

Would Not Involve a Significant 

Reduction in a Margin of Safety.

The proposed change does not adversely 

affect any plant safety limits, set points, or 

design parameters. The change also does not 

adversely affect the fuel, fuel cladding, 

Reactor Coolant System, or containment 

integrity. The RBACs will continue to be 

isolated during ES events. The modification 

eliminates significant waterhammers in the 

LPSW piping to the RBACs. 

The change will enhance the ability to 

provide LPSW flow to safety related loads 

following LOOP events. Therefore, the 

proposed change does not involve a 

reduction in a margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Anne W. 

Cottingham, Winston and Strawn LPP, 

1400 L Street, NW., Washington, DC 

20005. 

NRC Section Chief: Mary Jane RossLee, Acting. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–286, Indian Point 

Nuclear Generating Unit No. 3, 

Westchester County, New York

Date of amendment request: June 24, 

2004. 

Description of amendment request:

The proposed amendment would allow 

entry into a mode or other specified 

condition in the applicability of a 

Technical Specification (TS), while in a 

condition statement and the associated 

required actions of the TSs, 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 

would be revised to reflect the related 

changes to LCO 3.0.4, and Surveillance 

Requirement (SR) 3.0.4 would be 

revised to reflect the LCO 3.0.4 

allowance. 

This change was proposed by the 

industry’s TS 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 June 24, 2004. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below:

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

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: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc. 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Section Chief: Richard J. Laufer. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant, Oswego County, 

New York 

Date of amendment request:

September 1, 2004. 

Description of amendment request:

The proposed amendment would delete 

Technical Specification (TS) 5.6.1, 

‘‘Occupational Radiation Exposure 

Report,’’ and TS 5.6.4, ‘‘Monthly 

Operating Reports.’’

The NRC staff issued a notice of 

availability of a model no significant 

hazards consideration (NSHC) 

determination for referencing in license 

amendment applications in the Federal 

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

The licensee affirmed the applicability 

of the model NSHC determination in its 

application dated September 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 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57985

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: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Section Chief: Richard J. Laufer. 

Exelon Generation Company, LLC, and 

PSEG Nuclear LLC, Dockets Nos. 50–277 

and 50–278, Peach Bottom Atomic 

Power Station, Units 2 and 3, York and 

Lancaster Counties, Pennsylvania 

Date of application for amendments:

June 24, 2004. 

Description of amendment request:

The proposed amendment would 

incorporate several Technical 

Specification Task Force (TSTF) 

changes to the licensees Technical 

Specifications (TSs). The specific TSTF 

changes that would be incorporated are: 

(1) TSTF–5, Rev. 1, Delete Safety 

Limit Violation Notification 

Requirement—This change modifies TS 

Section 2.2 to remove the requirements 

to report safety limit violations. 

Associated references to Title 10 of the 

Code of Federal Regulations (10 CFR), 

Sections 50.72 and 50.73, are also 

removed. 

TSTF–208, Rev. 0, Extension of Time 

to Reach Mode 2 in LCO (Limiting 

Condition for Operation) 3.0.3—This 

TSTF modifies TS Section LCO 3.0.3 to 

revise the time to be in Mode 2 once 

LCO 3.0.3 is entered from 7 hours to a 

bracketed site-specific time depending 

on the individual plant’s ability to reach 

Mode 2 in a controlled shutdown. 

TSTF–222, Rev. 1, Control Rod Scram 

Time Testing and TSTF–229, Rev. 0, 

Revise Surveillance Requirement 3.2.2.2 

for Consistency with 3.1.4.4—This 

TSTF modifies the TSs to clarify the 

frequency of performing control rod 

scram time testing subsequent to 

performance of an outage that involved 

the movement of fuel. The current 

wording of Surveillance Requirement 

(SR) 3.1.4.1 could be interpreted that all 

control rods need to be scram time 

tested even if the shutdown was for a 

brief amount of time and only a limited 

amount of fuel was moved in the reactor 

(e.g., if only one bundle is moved in a 

mid-cycle fuel replacement). This 

change clarifies the intent of the TSs. 

TSTF–297, Rev. 1, and TSTF–227, 

Rev. 0—These two TSTFs affect the 

following three TS Sections:

3.3.2.2—Feedwater and Main Turbine 

High Water Level Trip 

Instrumentation 

3.3.4.1—Anticipated Transient Without 

Scram Recirculation Pump Trip 

(ATWS–RPT) Instrumentation 

3.3.4.2—End of Cycle Recirculation 

Pump Trip (EOC–RPT) 

Instrumentation

TSTF–297, Rev. 1—This TSTF 

modifies the TSs to add a new Required 

Action and corresponding note to allow 

affected feedwater pump(s) and main 

turbine valve(s) to be removed from 

service. This change is necessary to 

allow components to be removed from 

service to fulfill the safety function 

without a reduction in power to less 

than 25% rated thermal power. A 

similar note is added to TS Sections 

3.3.4.1 and 3.3.4.2 to provide the same 

clarification for when the associated 

Required Action is the appropriate 

action.

TSTF–227, Rev. 0—This TSTF 

modifies the TSs to eliminate ambiguity 

in the EOC–RPT Instrumentation 

Condition A. Since the LCO allows for 

having EOC–RPT instrumentation 

OPERABLE or certain fuel thermal 

limits are met, Condition A was 

inappropriately worded. The wording of 

Condition A is revised to add the word 

‘required’ if one or more channels are 

inoperable. Without the word ‘required’, 

one could interpret Condition A as 

needing entry even if the fuel thermal 

limits were being applied instead of 

applying the operability requirements to 

the EOC–RPT instrumentation. 

TSTF–295, Rev. 0, Post-Accident 

Monitoring Clarifications—This TSTF 

modifies the TSs to clarify that a 

separate Condition entry is allowed for 

each penetration flow path for the Post 

Accident Monitoring (PAM) 

instrumentation Primary Containment 

Isolation Valve (PCIV) indication 

function. 

TSTF–275, Rev. 0, ECCS 

Instrumentation Clarifications—This 

TSTF modifies the TSs to clarify which 

Emergency Core Cooling System (ECCS) 

instrumentation is required to be 

OPERABLE to support Emergency 

Diesel Generator (EDG) operability. 

Footnote (a) to Table 3.3.5.1–1 has been 

changed to only require the affected 

functions to be OPERABLE in Modes 4 

and 5 when the associated ECCS is 

required to be OPERABLE per LCO 

3.5.2. 

TSTF–306, Rev. 2, Traversing In-Core 

Probe Instrumentation Specification 

Requirements—This TSTF modifies the 

TSs by adding a note that penetration 

flow path may not be isolated 

intermittently under administrative 

control to conform to what is already 

allowed for similar specifications for 

Primary Containment Isolation Valves 

(PCIVs). Also, the Traversing In-core 

Probe (TIP) system isolation is set apart 

as a separate function including the 

allowance of isolating the penetration 

instead of requiring a plant shutdown. 

TSTF–416, Rev. 0, Clarification of 

LPCI Operability during Decay Heat 

Removal Operations—This TSTF 

modifies the TSs by moving the note 

that modifies Low Pressure Coolant 

Injection (LPCI) surveillances to the 

LCO in LCO 3.5.1 and LCO 3.5.2. These 

notes provide clarity that the LPCI may 

be considered OPERABLE during 

alignment and operation in the decay 

heat removal Mode. 

TSTF–17, Rev. 2, Containment 

Airlock Testing Frequency—This TSTF 

modifies the TSs to extend the testing 

frequency of the containment interlock 

mechanism from 184 days to 24 months. 

Also, the corresponding note for this 

surveillance is no longer required due to 

the longer surveillance frequency. 

57986 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

TSTF–30, Rev. 3, TSTF–323, Rev. 0, 

TSTF–45, Rev. 2, TSTF–46, Rev. 1, and 

TSTF–269, Rev. 2, Containment 

Isolation Valve Specification Changes—

These TSTFs modify TS Sections 3.6.1.3 

concerning Primary Containment 

Isolation Valves (PCIVs) and 3.6.4.2 

concerning Secondary Containment 

Isolation Valves (SCIVs). 

TSTF–30, Rev. 3 & TSTF–323, Rev. 

0—These TSTFs revise TS 3.6.1.3 to 

allow for a 72-hour completion time for 

a closed system flow path with an 

inoperable isolation valve and allow for 

a 72-hour completion time for a 

penetration flow path with an 

inoperable Excess Flow Check Valve 

(EFCV). 

TSTF–45, Rev. 2—This TSTF revises 

TSs 3.6.1.3 and 3.6.4.2 to revise 

surveillance requirements for valve lineups. Specifically, if a containment 

isolation valve is locked, sealed, or 

otherwise secured, they are not required 

to be verified to be closed during the 

performance of the surveillance test. 

TSTF–46, Rev. 1—This TSTF revises 

containment isolation valve 

surveillances to delete the reference to 

verifying the isolation time of ‘each 

power operated’ containment isolation 

valve and only require verification of 

each ‘automatic isolation valve’. 

TSTF–269, Rev. 2—This TSTF allows 

for verification of valve status by 

administrative means for repetitive 

verification of locked, sealed, or secured 

valves. 

TSTF–322, Rev. 2, Secondary 

Containment Operability Clarification—

This TSTF modifies the TSs to clarify 

the intent of the secondary containment 

boundary integrity. Associated 

surveillances currently imply that 

secondary containment would be 

inoperable if a Standby Gas Treatment 

(SGT) subsystem was inoperable. 

TSTF–276, Rev. 2, Power Factor for 

Emergency Diesel Generator (EDG) 

Surveillances—This TSTF modifies the 

TSs to allow for certain EDG testing to 

be performed even if the specified 

power factor cannot be achieved. 

TSTF–65, Rev. 1, Generic 

Organization Titles—This TSTF 

modifies the TSs to allow the use of 

generic organizational titles in place of 

plant-specific titles. Therefore, for the 

TSs, a change is requested to replace 

plant-specific titles with generic titles. 

TSTF–299, Rev. 0, Primary Coolant 

Sources Inspection Requirements—This 

TSTF modifies the TSs Section 5.2.2, 

‘Primary Coolant Sources Outside 

Containment’ to clarify the intent of 

refueling cycle intervals with respect to 

the system leak test requirements and 

adds a sentence that the leak test is 

subject to the provisions of Surveillance 

Requirements (SR) 3.0.2. 

TSTF–279, Rev. 0, Inservice Testing 

Program Clarifications—This TSTF 

modifies TSs Section 5.5.8, ‘‘Inservice 

Testing Program,’’ to delete the 

reference to ‘applicable supports’ as part 

of the description for the Inservice 

Testing Program. The applicable TS 

Section is 5.5.6. 

TSTF–118, Rev. 0, Diesel Generator 

Fuel Oil Testing Program 

Clarifications—This TSTF modifies TSs 

Section 5.5.13, ‘‘Diesel Fuel Oil Testing 

Program,’’ to allow for the provisions of 

SR 3.0.2 (25% extension) and SR 3.0.3 

(missed surveillance actions) to apply to 

surveillances. The applicable TS 

Section is 5.5.9. 

TSTF–106, Rev. 1, Diesel Generator 

Fuel Oil Testing Program 

Clarifications—This TSTF modifies the 

TSs to clarify that Section 5.5.10.b, 

concerning verification of the diesel fuel 

oil that was sampled meets the required 

ASTM properties, only applies to new 

fuel. As written, it could be interpreted 

that this testing is required for existing 

fuel that is routinely sampled. The 

applicable TS Section is 5.5.9.b. 

TSTF–152, Rev. 0, Routine Reporting 

Requirements Upgrade—This TSTF 

modifies the TSs to revise the 

Occupational Radiation Exposure 

Report and the Radioactive Effluent 

Release Report requirements to be 

consistent with other regulatory changes 

that have occurred. 

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. TSTF–5, Rev. 1, Delete Safety Limit 

Violation Notification Requirements. 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This action does not affect the plant or 

operation of the plant. The change simply 

removes duplicative information from the 

Technical Specifications that is covered in 

the NRC regulations. 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.

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do not challenge the 

performance or integrity of any safety related 

system. This change is considered an 

administrative action to remove duplicative 

reporting requirements. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

(3) Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

This administrative action does not involve 

any reduction in a margin of safety. Removal 

of duplicative information does not affect 

compliance with the regulations. Therefore, 

the proposed change does not involve a 

significant reduction in a margin of safety. 

B. TSTF–208, Rev. 0, Extension of Time to 

Reach Mode 2 in LCO 3.0.3. 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The time frame to take response action in 

accordance with LCO 3.0.3 is not an 

initiating condition for any accident 

previously evaluated and the accident 

analyses do not assume that any equipment 

is out of service such that LCO 3.0.3 is 

entered. The small increase in the time 

allowed to reach Mode 2 would not place the 

plant in any significantly increased 

probability of an accident occurring. The 

plant would already be proceeding to a plant 

shutdown condition because of the 1 hour 

requirement to initiate shutdown actions. 

There is no change in the time period to 

reach Mode 3. The Mode 3 Condition is the 

point where the plant is shutdown. 

Therefore, since there is no change to the 1 

hour requirement to initiate the shutdown 

nor any change to the time period to reach 

the shutdown Condition, the small change in 

the time to reach the Mode 2 status 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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. There are no plant physical 

alterations proposed. The proposed changes 

have no adverse effects on any safety-related 

system or component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The time period to reach Mode 3 and Mode 

4 are unaffected by this activity. This change 

simply provides a plant specific value for 

reaching Mode 2 if LCO 3.0.3 is entered 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57987

which is within the intent of LCO 3.0.3 for 

performing a controlled plant shutdown. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

C. TSTF–222. Rev. 1, Control, Red Scram 

Time Testing, and TSTF–229, Rev. 0, Revise 

Surveillance Requirement 3.2.2.2 for 

Consistency with 3.1.4.4 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

These changes are considered clarifications 

to the original intent of the Technical 

Specifications. Adequate testing of control 

rods is ensured by this change. Control rod 

operability is not affected by these changes. 

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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No.

This change is administrative in nature and 

does not affect any safety analyses 

assumptions. Adequate control rod testing 

continues to be maintained with 

implementation of this activity. Therefore, 

the proposed change does not involve a 

significant reduction in a margin of safety. 

(D) TSTF 297, Rev. 1, and TSTF 227, Rev. 

0, Enhancements to Feedwater/Main Turbine 

High Water Level Trip, EOC–RPT, and ATWS 

RPT Specifications 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

There are no changes to the plant 

configuration assumed for any accident. The 

removal from service of equipment that 

results in its safety function being met can 

not adversely affect the consequences of 

accidents previously evaluated. Other 

changes are administrative clarifications that 

have no affect on accidents. 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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The actions involved with this activity 

ensure that safety functions are met. There 

are no changes in the overall requirements of 

having trip instrumentation available for 

event mitigation. There are no affects on the 

plant safety analyses. Therefore, the 

proposed change does not involve a 

significant reduction in a margin of safety. 

(E) STF–295, Rev. 0, Post-Accident 

Monitoring Clarications 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The equipment involved with the revised 

Technical Specifications are for post-accident 

monitoring. This equipment has no 

possibility of increasing the probability of 

occurrence of the accident since it is 

monitoring equipment only. The 

consequences of an accident are not affected 

since this change maintains the original 

intent of the Technical Specifications in 

having available monitoring information for 

each PCIV penetration path. 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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The Technical Specifications continue to 

require appropriate post accident monitoring 

equipment to be OPERABLE. Adequate 

instrumentation for post-accident monitoring 

will be ensured by the Technical 

Specification requirements. There are no 

changes to the plant safety analyses involved 

with this change. Therefore, the proposed 

change does not involve a significant 

reduction in a margin of safety. 

(F) TSTF–275, Rev. 0, ECCS 

Instrumentation Clarifications 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The equipment involved is for mitigative 

purposes and will not affect the probability 

of occurrence of an accident. Technical 

Specifications ensures that adequate 

mitigative equipment continues to be 

OPERABLE for any event that may occur in 

Modes 4 and 5. This change is considered an 

upgrade to the specifications that will 

provide more consistency within the 

Technical Specifications. There are no 

changes to requirements that ensure 

appropriate Emergency Core Cooling Systems 

are OPERABLE. 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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

There is no impact on mitigative 

equipment that is required to respond to 

events while in Modes 4 and 5. There is no 

impact on the plant safety analyses. This 

change is considered as an upgrade to 

Technical Specifications that will improve 

consistency within the Technical 

Specifications. Therefore, the proposed 

change does not involve a significant 

reduction in a margin of safety. 

(G) TSTF–306, Rev. 2, Traversing In-Core 

Probe Instrumentation Specifications 

Requirements 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The addition of a note that the penetration 

flow path may be un-isolated under 

administrative control simply provides 

57988 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

consistency with what is already allowed 

elsewhere in [the] Technical Specifications. 

The isolation function of the TIP valves are 

mitigative equipment. They do not create any 

increased possibility of an accident since 

they are mitigative. Also, the operation of the 

manual shear valves is unaffected by this 

activity. The ability to manually isolate the 

TIP system by either the normal isolation 

valve or the shear valve would be unaffected 

by the inoperable instrumentation. Therefore, 

the same action as for manual isolation 

Functions provides an appropriate level of 

safety. 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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The addition of a note that the penetration 

flow path may be un-isolated under 

administrative control simply provides 

consistency with what is already allowed 

elsewhere in Technical Specifications. The 

ability to manually isolate the TIP system by 

either the normal isolation valve or the shear 

valve would be unaffected by the inoperable 

instrumentation. Therefore, the same action 

as for manual isolation Functions provides 

an appropriate level of safety. Therefore, the 

proposed change does not involve a 

significant reduction in a margin of safety. 

(H) TSTF–416, Rev. 0 Clarification of LPCI 

Operability during Decay Heat Removal 

Operations 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change makes the Technical 

Specifications and their Bases consistent in 

their consideration of an LPCI subsystem 

aligned for decay heat removal being 

considered OPERABLE for ECCS. The LCO 

3.5.1 and LCO 3.5.2 Bases state that a LPCI 

subsystem may be considered OPERABLE 

during alignment and operation for decay 

heat removal. As a result, no initiators to 

accidents previously evaluated are affected 

and no mitigating equipment assumed in the 

accidents previously evaluated are affected 

since the allowance for LPCI being 

considered operable during these type of 

shutdown cooling alignments or operations 

was the intent of the current technical 

Specifications. Consequently, the probability 

or consequences of an accident previous 

evaluated is not significantly increased. 

(2) Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change makes the Technical 

Specifications and their Bases consistent in 

their consideration of an LPCI subsystem 

aligned for decay heat removal being 

considered OPERABLE for ECCS. The LCO 

3.5.1 and LCO 3.5.2 Bases state that an LPCI 

subsystem may be considered OPERABLE 

during alignment and operation for decay 

heat removal. As the operability 

requirements of the LPCI subsystem are 

unaffected, the margin of safety is unaffected. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

(I) STF–17, Rev. 2, Containment Airlock 

Testing Frequency 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The containment airlock is considered as 

mitigative equipment. Therefore, there are no 

impacts on the probability of accidents. The 

proposed surveillance frequency assures that 

the interlock is working such that there is no 

unintentional opening of both airlock doors 

when containment is required. Because the 

interlock is assured to be working, there will 

be no significant increase in the 

consequences of an accident. There is no 

degradation in the ability of the interlock to 

assure the containment integrity function is 

maintained. 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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The frequency of 24 months for the 

interlock testing has been demonstrated to be 

adequate with regards to the reliability of the 

airlock. There is no impact on the leak testing 

requirements. There is no affect on the plant 

safety analyses. Therefore, the proposed 

change does not involve a significant 

reduction in a margin of safety. 

(J) TSTF–30, Rev. 3, TSTF–323, Rev. 0, 

TSTF–45, Rev. 2, TSTF–46, Rev. 1, and 

TSTF–269, Rev. 2, Containment Isolation on 

Valve Specification Changes 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The equipment affected by these changes is 

for mitigative purposes. Therefore, there 

cannot be an increase in the probability of 

occurrence of an accident. The controls 

required in the Technical Specifications are 

adequate to ensure that the containment 

barriers are ensured. Isolation valves will be 

assured to be in their correct positions. Also, 

inoperable isolation valves in closed systems 

and inoperable EFCVs have been evaluated to 

not have any significant impact to the 

consequences of an accident due to the 

closed system providing a barrier for the 

inoperable closed system isolation valve and 

bounding analyses have been performed for 

EFCV instrument line failures. 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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The equipment affected by these changes is 

for mitigative purposes. The controls 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57989

required in the Technical Specifications are 

adequate to ensure that the containment 

barriers are ensured. There is no affect on the 

plant safety analyses. Therefore, the 

proposed change does not involve a 

significant reduction in a margin of safety. 

(K) STF–322, Rev. 2, Secondary 

Containment Operability Clarification 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This change involves an administrative 

clarification to reflect the original intent of 

the Technical Specifications. There is no 

impact on the availability of the secondary 

containment. Additionally, secondary 

containment is mitigative equipment. 

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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

This change involves an administrative 

clarification to reflect the original intent of 

the Technical Specifications. There is no 

impact on the availability of the secondary 

containment. There is no impact on the plant 

safety analyses. Therefore, the proposed 

change does not involve a significant 

reduction in a margin of safety. 

(L) TSTF–276, Rev. 2, Power Factor for 

Emergency Diesel Generator (EDG) 

Surveillences 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

These changes only affect mitigative 

equipment and therefore, would not have an 

impact on the probability of an accident. 

Also, the performance of the surveillances 

ensures that mitigative equipment is capable 

of performing its intended function. 

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. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The performance of the surveillances 

ensures that mitigative equipment is capable 

of performing its intended function. There 

are no degradations in equipment readiness 

to mitigate design events. There is no adverse 

affect on the plant safety analysis. Therefore, 

the proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

(M) TSTF–65, Rev. 1, Generic 

Organizational Titles; 

TSTF–299, Rev. 0, Primary Coolant 

Sources Inspection Requirements; 

TSTF–279, Rev. 0, Inservice Testing 

Program Clarifications; 

TSTF–118, Rev. 0, and TSTF–106, Rev. 1, 

Diesel Generator Fuel Oil Testing Program 

Clarifications; 

TSTF–152, Rev. 0, Routine Reporting 

Requirement Upgrade 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The changes to Technical Specification 5.0, 

Administrative Controls, are considered 

administrative changes. There are no changes 

to plant structures, systems or components 

involved with this change. There are no 

degradations in the availability of mitigative 

plant equipment. The proposed changes 

provide enhancements to the administrative 

controls in Technical Specifications, 

therefore, there is no affect on any plant 

safety analyses; therefore, the proposed 

change does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

(2) Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

changes. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed changes have no adverse 

effects on any safety-related system or 

component and do 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 the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The changes to Technical Specification 5.0, 

Administrative Controls, are considered 

administrative changes. There are no changes 

to plant structures, systems or components 

involved with this change. There are no 

degradations in the availability of mitigative 

plant equipment. The proposed changes 

provide enhancements to the administrative 

controls in Technical Specifications; 

therefore, there is no affect on any plant 

safety analyses. Therefore, the proposed 

change does not involve a significant 

reduction in a margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for Licensee: Thomas S. 

O’Neill, Associate and General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Section Chief: Daniel S. Collins, 

Acting. 

FirstEnergy Nuclear Operating 

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

Ottawa County, Ohio 

Date of amendment request: August 

20, 2004. 

Description of amendment request:

The proposed amendment would revise 

the Technical Specifications (TS) 

regarding the requirement to 

demonstrate transfer of the unit A.C. 

electrical power supply to each offsite 

circuit and would increase the 

surveillance exceptions for the A.C. 

electrical sources in shutdown Modes 5 

and 6. Also, the proposed amendment 

would delete the TS requirement that 

the auto-connected loads to each 

emergency diesel generator (EDG) do 

not exceed the 2000-hour rating of the 

EDG.

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? 

57990 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

No. The proposed surveillance requirement 

changes do not alter the design or operation 

of any structure, system, or component. No 

previously analyzed accident scenario is 

changed. Initiating conditions and 

assumptions remain as previously analyzed. 

The revised surveillance requirements will 

continue to assure adequate performance of 

structures, systems, and components. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

(2) Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. The proposed surveillance requirement 

changes do not alter the design or operation 

of any structure, system, or component. No 

new or different accident initiators are 

created as a result of the proposed changes. 

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? 

No. The proposed surveillance requirement 

changes do not reduce or adversely affect the 

capabilities of the offsite and onsite electrical 

power sources. The revised surveillance 

requirements will continue to assure 

adequate performance of structures, systems, 

and components. The proposed changes do 

not affect conformance of the electrical 

power systems to the applicable design 

criteria. Therefore, the proposed changes do 

not involve a significant reduction in a 

margin of safety.

The 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: Mary E. 

O’Reilly, Attorney, FirstEnergy 

Corporation, 76 South Main Street, 

Akron, OH 44308. 

NRC Section Chief: Anthony J. 

Mendiola. 

Nuclear Management Company, LLC, 

Docket Nos. 50–282 and 50–306, Prairie 

Island Nuclear Generating Plant, Units 

1 and 2, Goodhue County, Minnesota 

Date of amendment request:

September 1, 2004. 

Description of amendment request:

The proposed amendments would 

revise the Operating Licenses’ licensing 

basis to allow use of the code for 

Generation of Thermal-Hydraulic 

Information for Containment, Version 

7.1patch1 (GOTHIC 7) to model Prairie 

Island Nuclear Generating Plant (PINGP) 

containment response for loss of coolant 

accidents (LOCA) and main steam line 

break (MSLB) accidents. The current 

PINGP containment response analyses 

are performed utilizing CONTEMPT. 

The Nuclear Management Company is 

making this request to support a 

transition option from internal analyses 

using CONTEMPT to an external 

analyses vendor (Westinghouse), which 

supports GOTHIC 7. 

Basis for proposed no significant 

hazards consideration determination:

As required by Title 10 of the Code of 

Federal Regulations (10 CFR), Part 

50.91(a), the licensee has provided its 

analysis of the issue of no significant 

hazards consideration, which is 

presented below:

(1) Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment will change the 

Prairie Island Nuclear Generating Plant 

licensing basis by allowing use of the 

Generation of Thermal-Hydraulic 

Information for Containment, Version 

7.1patch1, to model containment response 

for loss of coolant accident (LOCA) and main 

steam line break (MSLB) accidents. 

The containment is not an accident 

initiator, thus changing the containment 

modeling methodology does not increase the 

probability of an accident. This license 

amendment proposes to use a new 

methodology for modeling containment 

response analyses following an accident 

inside containment involving release of 

steam and water. This amendment does not 

alter the nuclear reactor core or reactor 

coolant system equipment, nor does it alter 

the methods or equipment used directly in 

mitigation of an accident. Thus radioactive 

releases inside containment due to an 

accident and radioactive releases from 

containment are not affected by the proposed 

change in analysis methodology. As 

discussed in Exhibits C and D, the Gothic 7 

sample results for the LOCA and MSLB 

transients predicted that the containment 

would remain below design pressure for both 

cases. Therefore, this change does not 

increase the consequences of an accident 

previously evaluated. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

(2) Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Response: No. 

The proposed amendment will change the 

Prairie Island Nuclear Generating Plant 

licensing basis by allowing use of the 

Generation of Thermal-Hydraulic 

Information for Containment, Version 

7.1patch1, to model containment response 

for LOCA and MSLB accidents. 

The proposed amendment does not involve 

changes to plant design, hardware, system 

operation, or procedures involved with 

containment function. The proposed changes 

include application of new methodology for 

analysis of containment response following a 

loss of coolant accident or steam line break 

accident. The results of the analyses are used 

to demonstrate that the acceptance criteria 

for the containment structure continue to be 

met. These changes do not create the 

possibility for a new or different kind of 

accident. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

(3) Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed amendment will change the 

Prairie Island Nuclear Generating Plant 

(PINGP) licensing basis by allowing use of 

the Generation of Thermal-Hydraulic 

Information for Containment, Version 

7.1patch1 (GOTHIC 7), to model containment 

response for LOCA and MSLB accidents.

The proposed licensing basis change to use 

GOTHIC 7 affects the design basis LOCA and 

MSLB containment accident analyses. As 

discussed in Exhibits C and D, the GOTHIC 

7 sample results for the LOCA and MSLB 

transients predicted that the containment 

would remain below design pressure for both 

cases. The GOTHIC 7 accuracy in this 

application has been verified through 

benchmark analyses against the current 

analyses of record, validated against 

recognized standard data, and found to be 

appropriate for application to the PINGP 

design basis accidents. Safety analysis 

acceptance criteria are satisfied and 

adherence to safety analysis acceptance 

criteria using GOTHIC 7 assures that 

Technical Specification limits will not be 

exceeded during normal operation. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: Jonathan Rogoff, 

Esquire, Vice President, Counsel & 

Secretary, Nuclear Management 

Company, LLC, 700 First Street, 

Hudson, WI 54016. 

NRC Section Chief: L. Raghavan. 

South Carolina Electric & Gas Company, 

South Carolina Public Service 

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

Summer Nuclear Station, Unit No. 1, 

Fairfield County, South Carolina 

Date of amendment request: May 21, 

2004. 

Description of amendment request:

The proposed amendment deletes the 

requirements from the technical 

specifications (TS) to maintain 

hydrogen recombiners and hydrogen 

monitors. Licensees were generally 

required to implement upgrades as 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57991

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 TS for nuclear power 

reactors currently licensed to operate. 

The revised 10 CFR 50.44, ‘‘Standards 

for Combustible Gas Control System in 

Light-Water-Cooled Power Reactors,’’

eliminated the requirements for 

hydrogen recombiners 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 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 May 21, 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 monitors 

are no longer required to mitigate designbasis accidents and, therefore, the hydrogen 

monitors do not meet the definition of a 

safety-related component as defined in 10 

CFR 50.2. Category 1 in RG 1.97 is intended 

for key variables that most directly indicate 

the accomplishment of a safety function for 

design-basis accident events. The hydrogen 

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. 

The regulatory requirements for the 

hydrogen 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, and 

removal of the hydrogen monitors from TS 

will not prevent an accident management 

strategy through the use of the severe 

accident management guidelines (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 the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, will not result in any failure mode 

not previously analyzed. The hydrogen 

recombiner and hydrogen monitor equipment 

was intended to mitigate a design-basis 

hydrogen release. The hydrogen recombiner 

and hydrogen 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, 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. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

Removal of hydrogen monitoring from TS 

will not result in a significant reduction in 

their functionality, reliability, and 

availability.

Based upon 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: Thomas G. 

Eppink, South Carolina Electric & Gas 

Company, Post Office Box 764, 

Columbia, South Carolina 29218. 

NRC Section Chief (Acting): Mary Jane 

Ross-Lee. 

Southern California Edison Company, et 

al., Docket Nos. 50–361 and 50–362, 

San Onofre Nuclear Generating Station, 

Units 2 and 3, San Diego County, 

California 

Date of amendment requests: August 

26, 2004. 

Description of amendment requests:

The proposed amendments would 

revise the Technical Specifications (TS) 

to implement ZIRLOTM fuel rod 

cladding material into the fuel design 

for San Onofre Nuclear Generating 

Station (SONGS), Units 2 and 3. 

Specifically, the licensee requests to add 

reference to ZIRLOTM clad fuel and filler 

rods in TS 4.2.1, ‘‘Fuel Assemblies,’’

and in TS 5.7.1.5, ‘‘Core Operating 

Limits Report (COLR),’’ add the 

following references to the list of 

analytical methods used to determine 

the core operating limits: ‘‘Calculative 

Methods for the C–E Nuclear Power 

Large Break LOCA [loss-of-coolant 

accident] Evaluation Model,’’ CENPD–1 

32, Supplement 4–P–A, August 2000, 

and ‘‘Implementation of ZIRLOTM

Cladding Material in CE [Combustion 

Engineering, Inc.] Nuclear Power Fuel 

Assembly Designs,’’ CENPD–404–P–A, 

November 2001.

57992 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change allows the use of 

methods required for the implementation of 

ZIRLOTM clad fuel rods in San Onofre 

Nuclear Generating Station (SONGS) Units 2 

and 3. The use of this methodology will not 

increase the probability of an accident 

because the plant systems will not be 

operated outside of design limits, no different 

equipment will be operated, and system 

interfaces will not change. 

As ZIRLOTM material is introduced to the 

reactor, transition cores will exist in which 

fuel assemblies containing ZIRLOTM and 

Zircaloy clad fuel rods are co-resident. Each 

type of fuel assembly (ZIRLOTM or Zircaloy 

clad fuel rods) will be evaluated based on the 

approved topical reports listed in TS 5.7.1.5. 

The use of this additional methodology 

will not increase the consequences of an 

accident because Limiting Conditions of 

Operation (LCOs) will continue to restrict 

operation to within the regions that provide 

acceptable results, and Reactor Protection 

System (RPS) trip setpoints will restrict plant 

transients so that the consequences of 

accidents will be acceptable. In addition, the 

consequences of the accidents will be 

calculated using NRC accepted 

methodologies. 

The transition cores that will exist as 

ZIRLOTM clad fuel is introduced to the 

reactor will not increase the consequences of 

an accident. Operation within the LCOs and 

RPS setpoints will continue to restrict plant 

transients so that the consequences of 

accidents will be acceptable. 

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 add any 

new equipment, modify any interfaces with 

any existing equipment, alter the 

equipment’s function, or change the method 

of operating the equipment. The proposed 

change does not alter plant conditions in a 

manner that could affect other plant 

components. The proposed change does not 

cause any existing equipment to become an 

accident initiator. The ZIRLOTM clad fuel rod 

design does not introduce features that could 

initiate an accident. 

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. 

Safety Limits ensure that Specified 

Acceptable Fuel Design Limits (SAFDLs) are 

not exceeded during steady state operation, 

normal operational transients and anticipated 

operational occurrences. All fuel limits and 

design criteria shall be met based on the 

approved methodologies defined in the 

topical reports. The RPS in combination with 

the LCOs will continue to prevent any 

anticipated combination of transient 

conditions for reactor coolant system 

temperature, pressure, and thermal power 

level that would result in a violation of the 

Safety Limits. Therefore, the proposed 

changes will have no impact on the margins 

as defined in the Technical Specification 

bases. 

The safety analyses determine the LCO 

settings and RPS setpoints that establish the 

initial conditions and trip setpoints, which 

ensure that the Design Basis Events 

(Postulated Accidents and Anticipated 

Operational Occurrences) analyzed in the 

Updated Final Safety Analysis Report 

(UFSAR) produce acceptable results. In 

addition, all fuel limits and design criteria 

shall be satisfied. The Design Basis Events 

that are impacted by the implementation of 

ZIRLOTM cladding will be analyzed using the 

NRC accepted methodology described in 

CENPD–404–P–A. 

The change in the fuel rod cladding 

material and the use of the Emergency Core 

Cooling System (ECCS) performance 

evaluation models, CENPD–132, Supplement 

4–P–A, ‘‘Calculative Methods for the CE 

Nuclear Power Large Break LOCA Evaluation 

Model’’ and CENPD–137, Supplement 2–P–

A, ‘‘Calculative Methods for the ABB [Asea 

Brown Boveri] CE Small Break LOCA 

Evaluation Model’’ will not involve a 

reduction in the margin of safety because 

LCOs and Limiting Safety System Settings 

(LSSS) will be adjusted, if necessary, to 

maintain acceptable results for the impacted 

Design Basis Events. 

Therefore, this change does not involve a 

significant reduction in a margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: Douglas K. 

Porter, Esquire, Southern California 

Edison Company, 2244 Walnut Grove 

Avenue, Rosemead, California 91770. 

NRC Section Chief: Robert Gramm. 

Southern Nuclear Operating Company, 

Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of Georgia, 

City of Dalton, Georgia, Docket Nos. 50–

321 and 50–366, Edwin I. Hatch Nuclear 

Plant, Units 1 and 2, Appling County, 

Georgia 

Date of amendment request: May 21, 

2004. 

Description of amendment request:

The proposed amendment would delete 

requirements from the Technical 

Specifications (TSs) to maintain 

hydrogen recombiners (Unit 2 only) and 

hydrogen and oxygen monitors. A 

notice of availability for this TS 

improvement using the consolidated 

line item improvement process was 

published in the Federal Register on 

September 25, 2003 (68 FR 55416). 

Licensees were generally required to 

implement upgrades as described in 

NUREG–0737, ‘‘Clarification of TMI 

[Three Mile Island] Action Plan 

Requirements,’’ and Regulatory Guide 

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, ‘‘Standards 

for Combustible Gas Control System in 

Light-Water-Cooled Power Reactors,’’

eliminated the requirements for 

hydrogen recombiners 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 

May 21, 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 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57993

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. Category 1 in RG 

1.97 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 the site 

survey monitoring that support modification 

of emergency plan protective action 

recommendations. 

Therefore, the elimination of the hydrogen 

recombiner requirements and 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 elimination of the hydrogen 

recombiner requirements and 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 recombiner and hydrogen and 

oxygen monitor equipment was intended to 

mitigate a design-basis hydrogen release. The 

hydrogen recombiner and 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 elimination of the hydrogen 

recombiner requirements and 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.

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, 

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 

revise Administrative Controls Section 

5.3.1 to replace the specific designation 

for the Health Physics Superintendent 

with a reference to the senior individual 

in charge of Health Physics, and to add 

flexibility to the qualification 

requirements for unit staff positions. 

This change supports Southern Nuclear 

Company’s ongoing initiative to achieve 

fleet standardization. 

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?

The proposed change to Technical 

Specifications Administrative Controls 

Section 5.3.1 involves the use of a more 

generic designation for the unit staff position 

responsible for Health Physics without 

reducing the level of authority required for 

that position. The proposed change also 

allows the flexibility to use an NRC 

accredited program for qualifying personnel 

to fill unit staff positions, which represents 

an acceptable alternative to the qualification 

requirements for these positions as currently 

specified in the Technical Specifications. 

Since the proposed changes are 

administrative in nature, they do not involve 

any physical changes to any structures, 

systems, or components, nor will their 

performance requirements be altered. The 

proposed changes also do not affect the 

operation, maintenance, or testing of the 

plant. Therefore, the response of the plant to 

previously analyzed accidents will not be 

affected. Consequently, the proposed changes 

do not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

(2) Does the proposed change create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

The proposed changes to the Technical 

Specifications will have no adverse impact 

on the overall qualification of the unit staff. 

The alternative use of an accredited program 

that has been endorsed by the NRC will 

ensure the educational requirements and 

power plant experience for each unit staff 

position are properly satisfied and will 

continue to fulfill applicable regulatory 

requirements. Also, since no change is being 

57994 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

made to the design, operation, maintenance, 

or testing of the plant, no new methods of 

operation or failure modes are introduced by 

the proposed changes. Therefore, the 

possibility of a new or different kind of 

accident from any previously evaluated is not 

created. 

(3) Does the proposed change involve a 

significant decrease in the margin of safety? 

The proposed changes to the Technical 

Specifications will have no adverse impact 

on the onsite organizational features 

necessary to assure safe operation of the 

plant. Lines of authority for plant operation 

are unaffected by the proposed changes. 

Also, the adoption of the more generic 

designation of the individual responsible for 

Health Physics will reduce the regulatory 

burden of having to devote limited resources 

to process a license amendment whenever a 

title change for this position is implemented. 

Accordingly, this reduction in regulatory 

burden and the option to use an accredited 

program endorsed by NRC to qualify the unit 

staff will improve plant efficiency without 

compromising plant safety. Therefore, the 

proposed changes do not involve a 

significant decrease in the margin of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ernest L. Blake, 

Jr., Esquire, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037. 

NRC Section Chief: Mary Jane RossLee, Acting. 

Southern Nuclear Operating Company, 

Inc, Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of amendment request: May 21, 

2004. 

Description of amendment request:

The proposed amendment would delete 

the requirements from the Technical 

Specifications (TS) to maintain 

hydrogen recombiners and hydrogen 

monitors. A notice of availability for 

this improvement using the 

consolidated line item improvement 

process was published in the Federal 

Register on September 25, 2003 (68 FR 

55416). 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, ‘‘Standards 

for Combustible Gas Control System in 

Light-Water-Cooled Power Reactors,’’

eliminated the requirements for 

hydrogen recombiners 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 determination 

(NSHC) 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 

May 21, 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 monitors 

are no longer required to mitigate designbasis accidents and, therefore, the hydrogen 

monitors do not meet the definition of a 

safety-related component as defined in 10 

CFR 50.2. Category 1 in RG 1.97 is intended 

for key variables that most directly indicate 

the accomplishment of a safety function for 

design-basis accident events. The hydrogen 

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. 

The regulatory requirements for the 

hydrogen 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, and 

removal of the hydrogen 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TSs, will not result in any failure mode 

not previously analyzed. The hydrogen 

recombiner and hydrogen monitor equipment 

was intended to mitigate a design-basis 

hydrogen release. The hydrogen recombiner 

and hydrogen 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, 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 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57995

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. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

Removal of hydrogen monitoring from TS 

will not result in a significant reduction in 

their functionality, reliability, and 

availability.

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: M. Stanford 

Blanton, Esq., Balch and Bingham, Post 

Office Box 306, 1710 Sixth Avenue 

North, Birmingham, Alabama 35201. 

NRC Section Chief: Mary Jane RossLee, Acting. 

Southern Nuclear Operating Company, 

Inc., et al., Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant, 

Units 1 and 2, Burke County, Georgia 

Date of amendment request: May 21, 

2004. 

Description of amendment request:

The proposed amendment would delete 

the requirements from the Technical 

Specifications (TSs) to maintain 

hydrogen recombiners and hydrogen 

monitors. A notice of availability for the 

TS improvement using the consolidated 

line item improvement process was 

published in the Federal Register on 

September 25, 2003 (68 FR 55416). 

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 LightWater-Cooled 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, ‘‘Standards 

for Combustible Gas Control System in 

Light-Water-Cooled Power Reactors,’’

eliminated the requirements for 

hydrogen recombiners 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 determination 

(NSHC) 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 

May 21, 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 monitors 

are no longer required to mitigate designbasis accidents and, therefore, the hydrogen 

monitors do not meet the definition of a 

safety-related component as defined in 10 

CFR 50.2. Category 1 in RG 1.97 is intended 

for key variables that most directly indicate 

the accomplishment of a safety function for 

design-basis accident events. The hydrogen 

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. 

The regulatory requirements for the 

hydrogen 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, and 

removal of the hydrogen 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 the site survey monitoring 

that support modification of emergency plan 

protective action recommendations.

Therefore, the elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TSs, will not result in any failure mode 

not previously analyzed. The hydrogen 

recombiner and hydrogen monitor equipment 

was intended to mitigate a design-basis 

hydrogen release. The hydrogen recombiner 

and hydrogen 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen 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. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

Removal of hydrogen monitoring from TSs 

will not result in a significant reduction in 

57996 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

their functionality, reliability, and 

availability.

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Arthur H. 

Domby, Troutman Sanders, 

NationsBank Plaza, Suite 5200, 600 

Peachtree Street, NE., Atlanta, Georgia 

30308–2216. 

NRC Section Chief: Mary Jane RossLee, Acting. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: August 

26, 2004. 

Description of amendment request:

The license amendment request 

proposes revising the Technical 

Specifications (TSs) to delete the TS 

requirements related to Hydrogen 

Analyzers and Hydrogen Recombiners 

consistent with NRC-approved TS Task 

Force (TSTF) Traveler number TSTF–

447, Revision 1, ‘‘Elimination of 

Hydrogen Recombiners and Change to 

Hydrogen and Oxygen Monitors.’’ The 

TS requirements related to Hydrogen 

Analyzers and Hydrogen Recombiners 

are contained in TS Tables 3.3–10 and 

4.3–10 and TSs 3.6.4.1 and 3.6.4.2. The 

availability of this TS improvement was 

announced in the Federal Register on 

September 25, 2003, as part of the 

Consolidated Line Item Improvement 

Process (CLIIP). 

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 analysis endorses the 

NRC staff’s generic no significant 

hazards consideration determination for 

TSTF–447 which was published in the 

Federal Register on September 25, 2003 

(68 FR 55416) as follows:

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 [Regulatory 

Guide] 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 

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. 

The regulatory requirements for the 

hydrogen 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, and 

removal of the hydrogen monitors from TS 

will not prevent an accident management 

strategy through the use of the SAMGs 

[Severe Accident Management Guidelines], 

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 the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, will not result in any failure mode 

not previously analyzed. The hydrogen 

recombiner and hydrogen monitor equipment 

was intended to mitigate a design-basis 

hydrogen release. The hydrogen recombiner 

and hydrogen 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 elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, 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. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

Removal of hydrogen monitoring from TS 

will not result in a significant reduction in 

their functionality, reliability, and 

availability.

The NRC staff proposes to determine 

that the request for amendments 

involves no significant hazards 

consideration. 

Attorney for licensee: A.H. Gutterman, 

Esq., Morgan, Lewis & Bockius, 1111 

Pennsylvania Avenue, NW., 

Washington, DC 20004. 

NRC Section Chief: Robert A. Gramm. 

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 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57997

Hazards Consideration Determination, 

and Opportunity for a Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see: (1) The applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (First Floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

Systems (ADAMS) Public Electronic 

Reading Room on the internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the NRC 

Public Document Room (PDR) Reference 

staff at 1–800–397–4209, (301) 415–

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

Calvert Cliffs Nuclear Power Plant, Inc., 

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

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of application for amendment:

August 6, 2002, as supplemented 

December 12, 2002, July 24, 2003, and 

March 1, May 20, and August 11, 2004. 

Brief description of amendment: The 

amendments replace the Technical 

Specifications 3.9.4 and 3.9.5 

requirements to close all containment 

penetrations providing direct access 

from the containment atmosphere to 

outside temperature with a set of more 

detailed and less restrictive 

requirements. 

Date of issuance: September 13, 2004. 

Effective date: As of the date of 

issuance to be implemented within 30 

days. 

Amendment No.: 268 and 244. 

Renewed Facility Operating License 

No. DPR–53: Amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: October 15, 2002 (67 FR 

63690). 

The December 12, 2002, July 24, 2003, 

March 1, 2004, and May 20, 2004, letters 

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 

August 11, 2004, letter withdrew the 

licensee’s requested changes to 

Technical Specification 3.9.3. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 13, 

2004. 

No significant hazards consideration 

comments received: No. 

Calvert Cliffs Nuclear Power Plant, Inc., 

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

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of application for amendments:

December 12, 2003. 

Brief description of amendments: The 

amendments delete Technical 

Specification Section 5.5.3, ‘‘PostAccident Sampling.’’

Date of issuance: September 15, 2004. 

Effective date: As of the date of 

issuance to be implemented within 30 

days. 

Amendment Nos.: 269 and 245. 

Renewed Facility Operating License 

Nos. DPR–53 and DPR–69: Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: April 13, 2004 (69 FR 19564). 

The Commission’s related evaluation 

of these amendments is contained in a 

Safety Evaluation dated September 15, 

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:

May 7, 2002, as supplemented April 7, 

2003 and July 19, 2004. 

Brief description of amendment: The 

amendment relocates the boration 

system Technical Specification (TS) 

requirements to the Technical 

Requirements Manual and the boron 

dilution analysis restrictions within the 

TSs. The amendment also revises the TS 

limiting condition for operation action 

and the surveillance requirements 

associated with the emergency core 

cooling, containment spray and cooling 

and auxiliary feedwater systems. 

Date of issuance: September 9, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 283. 

Facility Operating License No. DRP–

65: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: June 11, 2002 (67 FR 40021). 

The April 7, 2003, and July 19, 2004, 

supplements contained clarifying 

information and did not change the 

staff’s initial proposed finding of no 

significant hazards consideration. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 9, 

2004. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc., 

Docket No. 50–423, Millstone Power 

Station, Unit No. 3, New London 

County, Connecticut 

Date of application for amendment:

August 7, 2002, as supplemented 

November 5, 2003. 

Brief description of amendment: The 

amendment revises the Technical 

Specifications (TSs) related to safety 

system settings. Specifically, the 

amendment revises: (1) TS 1.0 

‘‘Definitions;’’ (2) TS 2.2.1 ‘‘Limiting 

Safety System Settings—Reactor Trip 

System Instrumentation Setpoints;’’ (3) 

TS 3.3.1 ‘‘Reactor Trip System 

Instrumentation;’’ (4) TS 3.3.2 

‘‘Engineered Safety Features Actuation 

System Instrumentation;’’ (5) TS 3.7.7 

‘‘Control Room Emergency Ventilation 

System;’’ and (6) TS 3.8.3.1 ‘‘Onsite 

Power Distribution—Operating.’’

Date of issuance: September 14, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 220. 

Facility Operating License No. DRP–

49: The amendment revised the TSs.

Date of initial notice in Federal 

Register: October 15, 2002 (67 FR 

63692). The November 5, 2003, 

supplement contained clarifying 

information and did not change the 

staff’s initial proposed finding of no 

significant hazards consideration. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 14, 

2004. 

No significant hazards consideration 

comments received: No. 

Duke Energy Corporation, Docket Nos. 

50–269, 50–270, and 50–287, Oconee 

Nuclear Station, Units 1, 2, and 3, 

Oconee County, South Carolina 

Date of application of amendments:

May 25, 2004. 

57998 Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 

Brief description of amendments: The 

amendments revised the licensing basis 

in the Updated Final Safety Analysis 

Report (UFSAR) to support installation 

of a low-pressure injection (LPI) cross 

connect inside containment. The 

changes to the UFSAR revise the 

licensing basis for selected portions of 

the core flood and LPI/Decay Heat 

Removal piping to allow exclusion of 

the dynamic effects associated with 

postulated rupture of that piping by 

application of leak-before-break 

technology. The amendments also revise 

the Technical Specifications (TSs) to 

delete TSs that will no longer apply 

when the LPI cross connect 

modification has been implemented. 

Date of issuance: September 2, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

during the fall 2004 refueling outage of 

Unit 3. 

Amendment Nos.: 340, 342, and 341. 

Renewed Facility Operating License 

Nos. DPR–38, DPR–47, and DPR–55:

Amendments revised the TSs. 

Date of initial notice in Federal 

Register: July 6, 2004 (69 FR 40673). 

The Commission’s related evaluation of 

the amendments is contained in a Safety 

Evaluation dated September 2, 2004. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket Nos. 50–282 and 50–306, Prairie 

Island Nuclear Generating Plant, Units 

1 and 2, Goodhue County, Minnesota 

Date of application for amendments:

January 20, 2004, as supplemented by 

letters dated May 19, July 13, and 

August 16, 2004. 

Brief description of amendments: The 

amendments change the Prairie Island 

technical specification (TS) on 

containment to implement a portion of 

TSs Task Force Traveler 5, ‘‘Revise 

containment requirements during 

handling irradiated fuel and core 

alterations.’’ The amendments also 

selectively implement an alternative 

source term per Title 10 of the Code of 

Federal Regulations, Section 50.67 to 

perform the radiological consequences 

analysis of the design-basis fuel 

handling accident which supports the 

proposed TS changes. 

Date of issuance: September 10, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 166 and 156. 

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: May 25, 2004 (69 FR 29769 ). 

The supplemental letters 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 10, 

2004.

No significant hazards consideration 

comments received: No. 

Pacific Gas and Electric Company, 

Docket No. 50–133, Humboldt Bay 

Power Plant, Unit 3, Humboldt County, 

California 

Date of application for amendment:

June 23, 2004. 

Brief description of amendment: The 

amendment removes a restriction from 

the Humboldt Bay Power Plant Unit 3 

license thereby permitting Pacific Gas 

and Electric to engage in active 

decommissioning of the facility. 

Date of issuance: September 10, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 35. 

Facility Operating License No. DPR–7:

This amendment revises the license. 

Date of initial notice in Federal 

Register: August 3, 2004 (69 FR 46587). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 10, 

2004. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of amendments request: June 28, 

2004, as supplemented by letter dated 

August 5, 2004. 

Brief Description of amendments: The 

amendments revise TS 3.4.13, ‘‘RCS 

[Reactor Coolant System] Operational 

Leakage,’’ TS 5.5.9, ‘‘Steam Generator 

[SG] Tube Surveillance Program,’’ and 

TS 5.6.10, ‘‘Steam Generator Tube 

Inspector Report.’’ They also add a new 

TS 3.4.17, ‘‘Steam Generator Tube 

Integrity.’’ These changes facilitate 

implementation of industry initiative 

NEI [Nuclear Energy Institute] 97–08, 

‘‘Steam Generator Program Guidelines,’’

which allows a comprehensive, 

performance-based approach to 

managing SG performance at Farley 

Nuclear Plant, Units 1 and 2. 

Date of issuance: September 10, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 163 and 156. 

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the 

Technical Specifications. 

Date of initial notice in Federal 

Register: August 3, 2004 (69 FR 46950). 

The supplemental letter dated August 5, 

2004, provided clarifying information 

that did not change the initial proposed 

no significant hazards consideration 

determinations. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 10, 

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 amendments:

September 6, 2002, as supplemented by 

letters dated December 19, 2002, March 

28, June 24, September 3, and October 

22, 2003. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications (TSs) to (1) relocate the 

pressure temperature limit curves and 

low temperature overpressure 

protection system limits to the Pressure 

and Temperature Limits Report (PTLR), 

(2) reference the PTLR in the affected 

TSs limiting conditions for operation 

and bases, including the addition of the 

PTLR to the definitions section of the 

TSs, and the addition of a new TS 

6.9.1.15 to the administrative controls 

section of the TSs, (3) relocate TS 

3.4.9.2, Pressurizer, to the Sequoyah 

Technical Requirements Manual and (4) 

revise TS 3.4.9.1, Pressure/Temperature 

Limits, Reactor Coolant System, and TS 

3.4.12, Low Temperature Over Pressure 

Protection Systems, to incorporate 

standard TSs requirements from 

NUREG–1431, Revision 2, ‘‘Standard 

Technical Specifications—

Westinghouse Plants.’’

Date of issuance: September 15, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 294 and 284. 

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: October 29, 2002 (67 FR 

66015). The supplemental letters 

provided clarifying information that did 

not expand the scope of the original 

application or change the initial 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Federal Register / Vol. 69, No. 187 / Tuesday, September 28, 2004 / Notices 57999

Safety Evaluation dated September 15, 

2004. 

No significant hazards consideration 

comments received: No.

Dated at Rockville, Maryland, this 17th day 

of September, 2004.

For the Nuclear Regulatory Commission. 

Ledyard B. Marsh, 

Director, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation.

[FR Doc. 04–21345 Filed 9–27–04; 8:45 am] 

BILLING CODE 7590–01–P

PRESIDENT’S COUNCIL ON 

INTEGRITY AND EFFICIENCY 

Senior Executive Service Performance 

Review Board Membership

AGENCY: Environmental Protection 

Agency (EPA).

ACTION: Notice.

SUMMARY: This notice sets forth the 

names and titles of the current 

membership of the PCIE Performance 

Review Board as of September 23, 2004.

EFFECTIVE DATE: September 28, 2004.

FOR FURTHER INFORMATION CONTACT:

Individual Offices of (the) Inspector 

General.

SUPPLEMENTARY INFORMATION: 

I. Background 

The Inspector General’s Act of 1978, 

as amended, has created independent 

audit and investigative units—Offices of 

(the) Inspector General—at 57 Federal 

agencies. In 1981, the President’s 

Council on Integrity and Efficiency 

(PCIE) was established by Executive 

Order as an interagency committee 

charged with promoting integrity and 

effectiveness in Federal programs. The 

PCIE is chaired by the Office of 

management and Budget’s Deputy 

Director for Management, and 

comprised principally of the 29 

Presidential appointed Inspectors 

General (IGs). The primary objectives of 

the PCIE are: (1) Mounting collaborative 

efforts to address integrity, economy, 

and effectiveness issues that transcend 

individual Federal agencies; and (2) 

increasing the professionalism and 

effectiveness of IG personnel throughout 

the Government. 

II. PCIE Performance Review Board 

Under 5 U.S.C. 4314(c)(1)–(5), and in 

accordance with regulations prescribed 

by the Office of Management and 

Budget, each agency is required to 

establish one or more Senior Executive 

Service (SES) performance review 

boards. The purpose of these boards is 

to review and evaluate the initial 

appraisal of a senior executive’s 

performance by the supervisor, along 

with any recommendations to the 

appointing authority relative to the 

performance of the senior executive. 

The current members of the President’s 

Council on Integrity and Efficiency 

Performance Review Board, as of 

September 23, 2004, were as follows:

Agency for International Development 

Phone Number: (202) 712–1170; PCIE/ECIE 

Liaison—Donna Rosa (202) 712–4993 

James R. Ebbitt—Deputy Inspector General 

Adrienne Rish—Assistant Inspector General 

for Investigation 

Robert S. Perkins—Counsel to the Inspector 

General 

Bruce Crandlemire—Assistant Inspection 

General for Audit 

Paula Hayes—Assistant Inspector General for 

Management 

DEPARTMENT OF AGRICULTURE 

Phone Number: (202) 720–8001 PCIE/ECIE 

Liaison—Cheryl Viani (202) 720–8001 

Joyce N. Fleischman—Deputy Inspector 

General 

Tracy A. LaPoint—Deputy Assistant 

Inspector General 

David R. Gray—Counsel to the Inspector 

General 

Suzanne M. Murrin—Assistant Inspector 

General for Policy Development and 

Resources Management 

Mark R. Woods—Assistant Inspector General 

for Investigations 

Jon E. Novak—Deputy Assistant Inspector 

General for Investigations 

Robert W. Young, Jr.—Assistant Inspector 

General for Audit 

Marlane T. Evans—Deputy Assistant 

Inspector General for Audit 

DEPARTMENT OF COMMERCE 

Phone Number: (202) 482–4661 PCIE/ECIE 

Liaison—Allison Lerner (202) 482–1577 

Edward L. Blansitt—Deputy Inspector 

General 

Anthony D. Mayo—Assistant Inspector 

General for Investigation 

Elizabeth T. Barlow—Counsel to the 

Inspector General 

Judith J. Gordon—Assistant Inspector 

General for Systems Evaluation 

Jill A. Gross—Assistant Inspector General for 

Inspections and Program Evaluation 

Jessica Rickenbach—Assistant Inspector 

General for Compliance and 

Administration 

DEPARTMENT OF DEFENSE 

Phone Number: (703) 604–8324 PCIE/ECIE 

Liaison—John R. Crane (703) 604–8324 

Charles W. Beardall—Director, Defense 

Criminal Investigative Service—Office of 

the Deputy Inspector General for 

Investigations 

Patricia Brannin—Assistant Inspector 

General for Audit Policy and Oversight, 

Office of the Deputy Inspector General for 

Inspections and Evaluations 

John R. Crane—Assistant Inspector General 

for Communications and Congressional 

Liaison 

Thomas Gimble—Deputy Inspector General 

for Intelligence 

Donald Horstman—Director, Investigations of 

Senior Officials, Office of the Deputy 

Inspector General for Investigations 

Francis E. Reardon—Deputy Inspector 

General for Auditing 

Mary Ugone—Assistant Inspector General, 

Acquisition Management, Office of the 

Deputy Inspector General for Auditing 

Keith West—Assistant Inspector General, 

Audit Followup and Technical Support, 

Office of the Deputy Inspector General for 

Auditing 

Daniel F. Willkens—Deputy Director, 

Defense Criminal Investigative Service, 

Office of the Deputy Inspector General for 

Investigations 

Shelton R. Young—Assistant Inspector 

General, Readiness and Logistics Support, 

Office of the Deputy Inspector General for 

Auditing 

DEPARTMENT OF EDUCATION 

Phone Number: (202) 205–6900 PCIE/ECIE 

Liaison—Kira Stankosky (202) 245–6997 

Thomas Carter—Deputy Inspector General 

Cathy Lewis—Assistant Inspector General for 

Evaluations, Inspections and Management 

Services 

Helen Lew—Assistant Inspector General for 

Audit Services 

George Rippey—Deputy Assistant Inspector 

General for Audit Services 

Thomas Sipes—Assistant Inspector General 

for Investigative Services 

Charles Coe—Assistant Inspector General for 

Information Technology and Computer 

Crimes Investigation 

Mary Mitchelson—Counsel to the Inspector 

General 

DEPARTMENT OF ENERGY 

Phone Number: (202) 586–4393 PCIE/ECIE 

Liaison—Arlene Acton (202) 586–1807 

John Hartman—Assistant Inspector General 

for Investigations 

Rickey Hass—Assistant Inspector General for 

Audit Operations 

Denise Smith—Assistant Inspector General 

for Resource Management 

Christopher Sharpley—Deputy Inspector 

General for Investigations and Inspections 

Linda Snider Director for Audit Policy and 

Administration Sanford Parnes Counsel to 

the Inspector General 

DEPARTMENT OF HEALTH AND HUMAN 

SERVICES 

Phone Number: (202) 619–3148 PCIE/ECIE 

Liaison—Sheri Denkensohn (202) 619–3148 

Lewis Morris—Chief Counsel to the Inspector 

General 

Tony Campbell—Assistant Inspector General 

for Operations Division, Office of 

Investigations 

Donald Dille—Acting Deputy Inspector 

General for Management and Policy 

Joe Green—Assistant Inspector General for 

Audit Management and Policy