76486 Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice 

Applications and Amendments to 

Facility Operating Licenses Involving 

No Significant Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from November 

25, 2004, through December 9, 2004. 

The last biweekly notice was published 

on December 7, 2004 (69 FR 70712). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. Within 60 days after the 

date of publication of this notice, the 

licensee may file a request for a hearing 

with respect to issuance of the 

amendment to the subject facility 

operating license and any person whose 

interest may be affected by this 

proceeding and who wishes to 

participate as a party in the proceeding 

must file a written request for a hearing 

and a petition for leave to intervene. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60-

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules and 

Directives Branch, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be delivered to Room 6D22, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

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

Copies of written comments received 

may be examined at the Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. The filing of 

requests for a hearing and petitions for 

leave to intervene is discussed below. 

Within 60 days after the date of 

publication of this notice, the licensee 

may file a request for a hearing with 

respect to issuance of the amendment to 

the subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

intervene. Requests for a hearing and a 

petition for leave to intervene shall be 

filed in accordance with the 

Commission’s ‘‘Rules of Practice for 

Domestic Licensing Proceedings’’ in 10 

CFR Part 2. Interested persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order.

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 76487

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/

requestor to relief. A petitioner/

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

A request for a hearing or a petition 

for leave to intervene must be filed by: 

(1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, Attention: Rulemaking and 

Adjudications Staff; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland, 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) E-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HearingDocket@nrc.gov; or (4) facsimile 

transmission addressed to the Office of 

the Secretary, U.S. Nuclear Regulatory 

Commission, Washington, DC, 

Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

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

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer of 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(I)–(viii). 

For further details with respect to this 

action, see the application for 

amendment which is available for 

public inspection at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the 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 PDR Reference staff at 1 (800) 397–

4209, (301) 415–4737 or by e-mail to 

pdr@nrc.gov.

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: June 10, 

2004.

Description of amendment request:

The amendments would revise 

Technical Specification (TS) 3.6.3, 

‘‘Containment Isolation Valves,’’ to 

allow the surveillance frequencies for 

leakage rate testing to be specified in the 

Catawba Nuclear Station Containment 

Leak Rate Testing Program. 

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:

First Standard 

Does the change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

No. 

This amendment will not change any 

previously evaluated accidents such as the 

postulated ‘‘Fuel Handling Accident (FHA) 

in Containment’’. No credit is assumed for 

VP containment isolation in the FHA within 

containment. The Containment Purge (VP) 

System and Hydrogen Purge (VY) System 

containment isolation valves are sealed 

closed during modes 1 through 4. The 

Containment Air Release and Addition (VQ) 

System containment isolation valves are 

designed to close within 5 seconds of a 

containment phase ‘‘A’’ isolation signal. The 

prevention and mitigation of these accidents 

is not affected by this change. 

Test data demonstrates that the likelihood 

of a malfunction of a resilient seal in one of 

the VP, VY, or VQ valves is not increased by 

this change in the surveillances. The systems 

will continue to be able to perform their 

design functions of isolating containment 

during the evaluated accidents. Test 

procedures will continue to monitor the 

leakage of these valves to ensure the design 

function will continue to be met. There is no 

impact on previously evaluated accidents 

since the valves will continue to close and 

seal or remain closed as originally assumed 

in the accident scenarios. 

Therefore, the changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Second Standard 

Does the change create the possibility of a 

new or different kind of accident from any 

accident previously evaluated? 

No. 

This change does not involve a physical 

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

different type of equipment will be installed) 

or a change in the methods governing any 

normal plant operation. The change does not 

alter assumptions made in the safety analyses 

or licensing basis. This change will not affect 

or degrade the ability of the Containment 

Purge System, Hydrogen Purge System, or 

Containment Air Release and Addition 

System valves to perform their specified 

safety functions. Therefore, the change does 

not create the possibility of a new or different 

kind of credible accident from any accident 

previously evaluated. 

Third Standard 

Does the proposed change involve a 

significant reduction in a margin of safety? 

No. 

SR 3.6.3.6 currently states: ‘‘The measured 

leakage rate for Containment Purge System 

and Hydrogen Purge System valves must be 

< 0.05 La (Design Leakage Rate) when 

pressurized to Pa (Design Containment 

Pressure). The measured leakage rate for 

Containment Air Release and Addition 

valves must be < 0.01 La when pressurized 

to Pa. These required maximum leak rates 

will not be changed by this amendment. 

Testing of these valves to measure leakage 

through the valve seats will continue, only at 

a different frequency based on past test 

results. This will be a nominal frequency of 

18 months for the VP System and in 

accordance with 10 CFR 50, Appendix J, 

Option B for the VQ and VY Systems. 

Therefore, the proposed changes listed above 

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. 

76488 Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 

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: John A. Nakoski. 

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: July 19, 

2004. 

Description of amendment request:

The amendments would revise 

Technical Specification (TS) 3.8.4, ‘‘DC 

Sources—Operating’’ and TS 3.8.6, 

‘‘Battery Cell Parameters’’ to allow for 

the replacement of the existing nickel 

cadmium diesel generator batteries with 

conventional lead acid batteries. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

(1) The proposed license amendments do 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The DG batteries are not accident initiating 

equipment; they are accident mitigating 

equipment. As such, they cannot affect the 

probability of any accident being initiated. 

The performance of the replacement batteries 

will exceed that of the existing batteries. 

Therefore, no accident consequences will be 

adversely impacted. 

(2) The proposed license amendments do 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

The DG batteries are not capable by 

themselves of initiating any accident. Other 

than the replacement of the batteries 

themselves and the associated modification 

work (e.g., installation of the battery HVAC 

system), no physical changes to the overall 

plant are being proposed. No changes to the 

overall manner in which the plant is 

operated are being proposed. Therefore, no 

potential for new accident types is generated. 

(3) The proposed license amendments do 

not involve a significant reduction in a 

margin of safety. 

Margin of safety is related to the 

confidence in the ability of the fission 

product barriers to perform their intended 

functions. These barriers include the fuel 

cladding, the reactor coolant system, and the 

containment. The modification to replace the 

DG batteries will not have any impact on 

these barriers. In addition, no accident 

mitigating equipment will be adversely 

impacted as a result of the battery 

replacement. The replacement batteries will 

have overall performance capabilities equal 

to or greater than those for the existing 

batteries. Therefore, existing safety margins 

will be preserved.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration.

Attorney for licensee: Ms. Lisa F. 

Vaughn, Legal Department (PB05E), 

Duke Energy Corporation, 422 South 

Church Street, Charlotte, North Carolina 

28201–1006. 

NRC Section Chief: John A. Nakoski. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of amendment request:

September 21, 2004. 

Description of amendments request:

The proposed amendment would delete 

the requirements from the technical 

specifications (TS) to maintain 

hydrogen recombiners and hydrogen 

and oxygen monitors. Licensees were 

generally required to implement 

upgrades as described in NUREG–0737, 

‘‘Clarification of TMI [Three Mile 

Island] Action Plan Requirements,’’ and 

Regulatory Guide (RG) 1.97, 

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

Plant and Environs Conditions During 

and Following an Accident.’’

Implementation of these upgrades was 

an outcome of the lessons learned from 

the accident that occurred at TMI Unit 

2. Requirements related to combustible 

gas control were imposed by Order for 

many facilities and were added to or 

included in the TS for nuclear power 

reactors currently licensed to operate. 

The revised 10 CFR 50.44, ‘‘Combustible 

gas control for nuclear 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 no significant hazards 

consideration determination in its 

application dated September 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 and 

oxygen monitors are no longer required to 

mitigate design-basis accidents and, 

therefore, the hydrogen monitors do not meet 

the definition of a safety-related component 

as defined in 10 CFR 50.2. RG 1.97, Category 

1, is intended for key variables that most 

directly indicate the accomplishment of a 

safety function for design-basis accident 

events. The hydrogen and oxygen monitors 

no longer meet the definition of Category 1 

in RG 1.97. As part of the rulemaking to 

revise 10 CFR 50.44, the Commission found 

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

appropriate categorization for the hydrogen 

monitors because the monitors are required 

to diagnose the course of beyond design-basis 

accidents. Also, as part of the rulemaking to 

revise 10 CFR 50.44, the Commission found 

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

appropriate categorization for the oxygen 

monitors, because the monitors are required 

to verify the status of the inert containment. 

The regulatory requirements for the 

hydrogen and oxygen monitors can be 

relaxed without degrading the plant 

emergency response. The emergency 

response, in this sense, refers to the 

methodologies used in ascertaining the 

condition of the reactor core, mitigating the 

consequences of an accident, assessing and 

projecting offsite releases of radioactivity, 

and establishing protective action 

recommendations to be communicated to 

offsite authorities. Classification of the 

hydrogen monitors as Category 3, 

classification of the oxygen monitors as 

Category 2 and removal of the hydrogen and 

oxygen monitors from 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 

(EOPs), 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 and oxygen monitor 

requirements, including removal of these 

requirements from TS, does not involve a 

Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 76489

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 TS, 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 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. 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 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: Thomas C. 

Poindexter, Esq., Winston & Strawn, 

1400 L Street, NW., Washington, DC 

20005–3502. 

NRC Section Chief: Robert A. Gramm. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of amendment request: April 14, 

2004. 

Description of amendment request:

The proposed amendment would add a 

new section to the Technical 

Specifications (TSs) and two new 

Limiting Conditions for Operations 

(LCOs) to allow certain reactor coolant 

system (RCS) hydrostatic and system 

leakage pressure tests to be performed 

with the reactor pressure vessel 

temperature above 212° Fahrenheit (F). 

The first LCO would allow specified TS 

requirements to be changed to permit 

performance of special tests and 

operations, which otherwise could not 

be performed if required to comply with 

the requirements of the TSs. The second 

LCO would require reactor low water 

level instrumentation, standby gas 

treatment system, and secondary 

containment to be OPERABLE to allow 

certain RCS pressure tests to be 

performed with the reactor pressure 

vessel temperature above 212° F, and 

provides for an exemption from the 

requirements for OPERABILITY for 

other systems that currently go into 

effect when in Hot Shutdown or when 

RCS temperature is greater than 212° F. 

It will also update the Table of Contents 

to reflect the proposed changes. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration. The Nuclear Regulatory 

Commission (NRC) staff has reviewed 

the licensee’s analysis against the 

standards of 10 CFR 50.92(c). The NRC 

staff’s review is presented below. 

1. Does the change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

The proposed change will not involve 

a significant increase in the probability 

or consequences of an accident 

previously evaluated. The probability of 

an accident previously evaluated is not 

significantly increased because the 

proposed change will not alter the 

method by which RCS hydrostatic 

pressure and leak testing is performed. 

Under this proposed change the 

secondary containment, standby gas 

treatment system and associated 

initiation instrumentation are required 

to be operable during the performance 

of RCS hydrostatic pressure and leak 

testing and would be capable of 

handling any airborne radioactivity or 

steam leaks that could occur. The 

required pressure testing conditions 

provide adequate assurance that the 

consequences of a steam leak will be 

conservatively bounded by the 

consequences of a main steamline break 

(MSLB) outside the primary 

containment. Accordingly, the 

consequences of previously evaluated 

accidents are not increased 

significantly. 

The proposed update to the Table of 

Contents is editorial in nature. Since 

this update is administrative in nature, 

it cannot increase the probability or 

consequences of a previously analyzed 

accident. 

Therefore, the proposed change does 

not involve a significant increase in the 

probability or consequences of an 

accident previously evaluated. 

2. Does the proposed change create 

the possibility of a new or different kind 

of accident from any previously 

evaluated? 

The proposed amendment change will 

not alter the way that hydrostatic 

pressure and leak testing is performed. 

Therefore, the proposed change will not 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated. 

3. Does the proposed change involve 

a significant reduction in a margin of 

safety? 

The proposed amendment will not 

involve a significant reduction in a 

margin of safety for a postulated MSLB 

outside of primary containment. The 

proposed changes and additions result 

in increased system operability 

requirements above those that currently 

exist during the performance of RCS 

hydrostatic pressure and leak testing. 

The incremental increase in stored 

energy in the vessel during testing will 

be conservatively bounded by the 

consequences of the postulated MSLB 

outside of primary containment. 

Therefore, the proposed change does not 

involve a significant reduction in a 

margin of safety. 

The proposed update to the Table of 

Contents is editorial in nature. Since 

this update is administrative in nature, 

the proposed change does not involve a 

significant reduction in a margin of 

safety 

Based on this review, it appears that 

the three standards of 10 CFR 50.92(c)) 

are satisfied. Therefore, the NRC staff 

proposes to determine that the 

76490 Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. M. Fulton, 

Esquire, Assistant General Counsel, 

Pilgrim Nuclear Power Station, 600 

Rocky Hill Road, Plymouth, 

Massachusetts, 02360–5599. 

NRC Section Chief: Darrell Roberts. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of amendment request:

September 2, 2004. 

Description of amendment request:

The proposed amendment would revise 

Technical Specification 4.5.B.2.2 (TS) to 

change the surveillance requirement 

frequency for air testing the drywell and 

suppression pool (torus) spray headers 

and nozzles from ‘‘once every 5 years’’

to ‘‘following maintenance that could 

result in nozzle blockage.’’

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, 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 drywell and torus headers and spray 

nozzles are not assumed to be initiators of 

any accidents previously evaluated. 

Maintenance practices and normal 

environmental conditions to which the 

system is subjected are adequate to ensure 

operability of the systems. Since the system 

will be able to perform its accident mitigation 

function, the consequences of accident 

previously evaluated are not increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident [from any accident] previously 

[evaluated]? 

Response: No. 

The revised surveillance does not 

introduce any new mode of plant operation, 

does not involve physical modification of the 

plant, or any new operating modes, and 

cannot introduce new accident initiators. 

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. 

Maintenance practices and normal 

environmental conditions to which the 

system is subjected are adequate to ensure 

operability of the systems. As the spray 

nozzles are expected to remain fully capable 

of performing their post-accident mitigation 

function, margin of safety is not reduced. 

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 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: J. M. Fulton, 

Esquire, Assistant General Counsel, 

Pilgrim Nuclear Power Station, 600 

Rocky Hill Road, Plymouth, 

Massachusetts, 02360–5599. 

NRC Section Chief: Darrell J. Roberts. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of amendment request:

September 2, 2004. 

Description of amendment request:

The proposed amendment would 

remove a license condition that 

currently requires the reactor not to be 

operated for more than 24 hours if one 

recirculation loop is out of service. It 

would revise Technical Specifications 

(TSs) to allow the minimum critical 

power ratio (MCPR) safety limit to be 

changed for single loop operations 

(SLOs). It would also revise the current 

jet pump limiting condition for 

operation and surveillance requirements 

to allow for the conduct of a TS required 

surveillance during SLOs. The proposed 

amendment would modify the TSs to 

address SLO operating conditions and 

restrictions, and delete a TS condition 

related to thermal-hydraulic stability. It 

would update the TSs for average planar 

linear heat generation rate for SLOs, and 

update the thermal power applicability 

restrictions to be consistent with 

NUREG–1433, Revision 3, ‘‘Standard 

Technical Specifications for General 

Electric Boiling-Water Reactors.’’ It 

would also revise the TSs for linear heat 

generation rate and MCPR for thermal 

power applicability restrictions. The 

proposed amendment makes an 

administrative change to have MCPR 

recalculated when reactor power is 

equal to or greater than 25 percent. 

Lastly, it would update the TSs’ table of 

contents and TS pages to 

administratively reflect all of these 

proposed changes.

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 license and technical 

specification changes will allow the plant to 

be operated with one recirculation pump for 

longer than 24 hours provided that 

appropriate limits are instituted. Extended 

single recirculation loop operation has been 

evaluated and methodologies have been 

established for determining appropriate 

operating limits. Implementation of the 

single recirculation loop operating limits 

ensures that system operation is in 

conformance with the conditions established 

to minimize the probability of accidents and 

the associated consequences. Required 

completion times for implementing the 

system operating limits and restoring out of 

specification limits minimize the probability 

that an accident occurs when out of 

specification conditions exist while allowing 

for deliberate operator action. 

Therefore, this proposed amendment does 

not involve a significant increase in the 

probability of occurrence 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 license and technical 

specification changes will allow plant 

operation with a single recirculation loop for 

longer than 24 hours. The proposed changes 

introduce an additional recirculation systemoperating mode, however, existing system 

component operating equipment or operating 

characteristics will not change. The Pilgrim 

Station Single Loop Analysis Report 

identifies required operating limits that apply 

when the system will be operated in the 

single loop operation mode. Implementation 

of these operating limits will ensure that the 

system is operated in accordance with 

design. Additionally, revised jet pump 

surveillance ensures that loop specific 

surveillance is performed as required to 

validate the bounding assumptions of 

existing accident analyses. As such, no new 

failure mechanisms are created and existing 

design evaluations bound system operation. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident. 

3. Does the proposed change involve a 

significant reduction in [a] margin of safety? 

Response: No. 

The proposed license and technical 

specification changes identify the operating 

limits that apply to single recirculation loop 

operation. These proposed recirculation 

system limits were identified to ensure that 

system operation would be in conformance to 

the conditions evaluated in applicable 

accident and transient analyses. 

Implementation of the proposed limits for 

single recirculation loop operation ensures 

that safety margins are maintained. Required 

completion times for implementing the 

system operating limits minimizes the 

Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 76491

possibility that an accident occurs when out 

of specification conditions exist. 

Therefore, the proposed changes do not 

involve a significant reduction in [a] margin 

of safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. M. Fulton, 

Esquire, Assistant General Counsel, 

Pilgrim Nuclear Power Station, 600 

Rocky Hill Road, Plymouth, 

Massachusetts 02360–5599. 

NRC Section Chief: Darrell Roberts. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc. 

(licensee), Docket No. 50–271, Vermont 

Yankee Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request:

September 16, 2003 as supplemented by 

letter dated March 15, 2004. 

Description of amendment request:

The proposed amendment would 

relocate the current definition of 

surveillance frequency to new Technical 

Specification (TS) Sections 4.0.2 and 

4.0.3, and revise the requirements for 

missed surveillance in Section 4.0.3. 

This change is consistent with NRCapproved Industry/Technical 

Specification Task Force (TSTF) change 

TSTF–358, Revision 5. The proposed 

change would allow a longer period of 

time to perform a missed surveillance. 

The time is extended from the current 

limit of up to 24 hours or up to the limit 

of the specified frequency, whichever is 

less; to up to 24 hours or up to the limit 

of the specified frequency, whichever is 

greater. In conjunction with the 

proposed change, the proposed 

amendment would add the 

requirements for a Bases Control 

Program which is consistent with 

Section 5.5 of NUREG 1433. In addition, 

the current definition of surveillance 

interval (definition ‘‘Z’’) would be reworded and relocated to new Section 

4.0.1 consistent with Surveillance 

Requirement 3.0.1 of NUREG 1433. 

Appropriate Bases, also consistent with 

NUREG 1433 would be adopted for the 

new sections. An editorial change 

would be made to TS 6.7.C which 

references the current definition of 

surveillance frequency to now reference 

the new Section 4.0.2. 

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on June 14, 2001 (66 FR 32400), 

on possible amendments concerning 

missed surveillances, 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 September 28, 2001 (66 FR 

49714). The licensee affirmed the 

applicability of the model NSHC 

determination in its application dated 

September 16, 2003. The model NSHC 

determination analysis for changes to 

the TS associated with missed 

surveillances, and the NSHC 

determination analysis provided by the 

licensee for the remaining TS changes, 

is provided herein. 

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 

With regard to the proposed change to the 

TS associated with missed surveillances, the 

proposed change relaxes the time allowed to 

perform a missed surveillance. The time 

between surveillances is not an initiator of 

any accident previously evaluated. 

Consequently, the probability of an accident 

previously evaluated is not significantly 

increased. The equipment being tested is still 

required to be operable and capable of 

performing the accident mitigation functions 

assumed in the accident analysis. As a result, 

the consequences of any accident previously 

evaluated are not significantly affected. Any 

reduction in confidence that a standby 

system might fail to perform its safety 

function due to a missed surveillance is 

small and would not, in the absence of other 

unrelated failures, lead to an increase in 

consequences beyond those estimated by 

existing analyses. The addition of a 

requirement to assess and manage the risk 

introduced by the missed surveillance will 

further minimize possible concerns. 

Therefore, this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

With regard to the remaining proposed 

changes to the TSs, the proposed changes do 

not involve physical changes to the plant or 

introduce any new modes of operation. 

Accordingly, continued assurance is 

provided that the process variables, 

structures, systems, and components are 

maintained such that there will be no 

degradation of any fission product barrier 

which could increase the radiological 

consequences of an accident. Therefore, the 

proposed changes do 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 

With regard to the proposed changes to the 

TSs associated with missed surveillances, the 

proposed change does not involve a physical 

alteration of the plant (no new or different 

type of equipment will be installed) or a 

change in the methods governing normal 

plant operation. A missed surveillance will 

not, in and of itself, introduce new failure 

modes or effects and any increased chance 

that a standby system might fail to perform 

its safety function due to a missed 

surveillance would not, in the absence of 

other unrelated failures, lead to an accident 

beyond those previously evaluated. The 

addition of a requirement to assess and 

manage the risk introduced by the missed 

surveillance will further minimize possible 

concerns. Thus, this change does not create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

With regard to the remaining proposed 

changes to the TSs, the proposed changes do 

not involve a physical alteration of the plant 

(no new or different type of equipment will 

be installed) or a change in the methods 

governing normal plant operation. Thus, the 

changes do 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 the Margin 

of Safety 

With regard to the proposed changes to the 

TSs associated with missed surveillances, the 

extended time allowed to perform a missed 

surveillance does not result in a significant 

reduction in the margin of safety. As 

supported by the historical data, the likely 

outcome of any surveillance is verification 

that the LCO [limiting condition for 

operation] is met. Failure to perform a 

surveillance within the prescribed frequency 

does not cause equipment to become 

inoperable. The only effect of the additional 

time allowed to perform a missed 

surveillance on the margin of safety is the 

extension of the time until inoperable 

equipment is discovered to be inoperable by 

the missed surveillance. However, given the 

rare occurrence of inoperable equipment, and 

the rare occurrence of a missed surveillance, 

a missed surveillance on inoperable 

equipment would be very unlikely. This 

must be balanced against the real risk of 

manipulating the plant equipment or 

condition to perform the missed surveillance. 

In addition, parallel trains and alternate 

equipment are typically available to perform 

the safety function of the equipment not 

tested. Thus, there is confidence that the 

equipment can perform its assumed safety 

function. Therefore, these changes do not 

involve a significant reduction in a margin of 

safety. 

With regard to the remaining proposed 

changes to the TSs, the administrative 

changes do not alter the basic operation of 

process variables, systems, or components as 

described in the safety analysis. No new 

equipment is introduced. Accordingly, the 

76492 Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 

proposed changes do not involve a 

significant reduction in a margin of safety.

Therefore, this change does not 

involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and its endorsement 

of the model NSHC for missed 

surveillances and, based on this review, 

it appears that the three standards of 10 

CFR 50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. David R. 

Lewis, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037–1128. 

NRC Section Chief: Allen G. Howe. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request: October 

5, 2004. 

Description of amendment request:

The proposed amendment would revise 

Technical Specification Section 6.7.C 

‘‘Primary Containment Leak Rate 

Testing Program,’’ to allow a one-time 

extension to the 10-year interval for 

performing the next Type A 

containment integrated leak rate test 

(ILRT). Specifically, the change would 

allow the test to be performed within 15 

years from the last ILRT, which was 

performed in April 1995. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below:

1. The operation of Vermont Yankee 

Nuclear Power Station in accordance with 

the proposed amendment will not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The proposed revision to Technical 

Specifications adds a one-time extension to 

the current interval for Type A testing. The 

current test interval of 10.6 years, based on 

past performance, is extended on a one-time 

basis to fifteen years from the last Type A 

test. The proposed extension to Type A 

testing cannot increase the probability of an 

accident previously evaluated since the 

containment Type A testing extension is not 

a modification and the test extension is not 

of a type that could lead to equipment failure 

or accident initiation. 

The proposed extension to Type A testing 

does not involve a significant increase in the 

consequences of an accident since research 

documented in NUREG–1493 has found that, 

generically, very few potential containment 

leakage paths are not identified by Type B 

and C tests. The NUREG concluded that 

reducing the Type A (ILRT) testing frequency 

to once per twenty years was found to lead 

to an imperceptible increase in risk. These 

generic conclusions were confirmed by a 

plant specific risk analysis performed using 

the current Vermont Yankee Probabilistic 

Safety Assessment (PSA) internal events 

model that concluded the consequences are 

low to negligible. 

Testing and inspection programs in place 

also provide a high degree of assurance that 

the containment will not degrade in a 

manner detectable only by Type A testing. 

The last two successful Type A tests indicate 

a very leak tight containment. Type B and C 

testing required by Technical Specifications 

will identify any containment opening such 

as valves that would otherwise be detected 

by the Type A tests. Inspections, including 

those required by the ASME [American 

Society of Mechanical Engineers] code and 

the Maintenance Rule are performed in order 

to identify indications of containment 

degradation that could affect that leak 

tightness. 

Therefore, the proposed changes do not 

represent a significant increase in the 

probability or consequences of an accident 

previously analyzed. 

2. The operation of Vermont Yankee 

Nuclear Power Station in accordance with 

the proposed amendment will not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed revision to Technical 

Specifications adds a one time extension to 

the current interval for Type A testing. The 

current test interval of 10.6 years, based on 

past performance, would be extended on a 

one time basis to fifteen years from the last 

Type A test. The proposed extension to Type 

A testing cannot create the possibility of a 

new or different type of accident since there 

are no physical changes being made to the 

plant and there are no changes to the 

operation of the plant that could introduce a 

new failure mode creating an accident or 

affecting the mitigation of an accident. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously analyzed. 

3. The operation of Vermont Yankee 

Nuclear Power Station in accordance with 

the proposed amendment will not involve a 

significant reduction in a margin of safety. 

The proposed revision to Technical 

Specifications adds a one time extension to 

the current interval for Type A testing. The 

current test interval of 10.6 years, based on 

past performance, would be extended on a 

one time basis to fifteen years from the last 

Type A test. The proposed extension to Type 

A testing will not significantly reduce the 

margin of safety. The NUREG–1493 generic 

study of the effects of extending containment 

leakage testing found that a 20-year extension 

in Type A leakage testing resulted in an 

imperceptible increase in risk to the public. 

NUREG–1493 found that, generically, the 

design containment leakage rate contributes 

about 0.1 percent to the individual risk and 

that the decrease in Type A testing frequency 

would have a minimal affect on this risk 

since 95% of the potential leakage paths are 

detected by Type C testing. This was further 

confirmed by a plant specific risk assessment 

using the current Vermont Yankee PSA 

internal events model that concluded the risk 

associated with this change is negligibly 

small and/or non-risk significant. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety.

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. David R. 

Lewis, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037–1128.

NRC Section Chief: Allen G. Howe. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request: October 

6, 2004. 

Description of amendment request:

The proposed amendment would revise 

Technical Specification Surveillance 

Requirement 4.5.B.1 related to air 

testing of the drywell spray headers and 

nozzles. Specifically, the amendment 

would change the test frequency from 

once every 5 years to following 

maintenance that could result in nozzle 

blockage. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration. The NRC staff has

Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 76493

reviewed the licensee’s analysis against 

the three standards of 10 CFR 50.92(c). 

The NRC staff’s analysis 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 amendment would 

revise the Technical Specification 

surveillance requirements associated 

with the air test of the drywell spray 

headers and nozzles. The frequency of 

the air test would be changed from a 

fixed 5-year frequency to following 

maintenance that could result in nozzle 

blockage. 

This surveillance test is performed 

while the plant is in a cold shutdown 

condition and the equipment is not 

required to be operable. The testing is to 

verify that the spray headers and 

nozzles are not obstructed. The 

proposed change in the surveillance test 

frequency will not result in any design 

changes to systems, structures, or 

components, or their method of 

operation. The drywell spray headers 

and nozzles are not initiators of any 

accidents previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability of any accident previously 

evaluated. 

The drywell spray headers provide a 

means to control both temperature and 

pressure inside the primary 

containment, within design limits, 

under post-accident conditions. Due to 

the system design and operation 

considerations discussed in the 

licensee’s application, the potential for 

corrosion product formation is 

minimized. In addition, the Vermont 

Yankee foreign material exclusion 

program has been judged to be sufficient 

to ensure that foreign material is not 

inadvertently introduced into the 

system. The proposed testing 

requirements are considered sufficient 

to provide a high degree of confidence 

that containment spray will function 

when required. Therefore, the proposed 

change does not involve a significant 

increase in the consequences of any 

accident previously evaluated. 

2. Does the proposed change create 

the possibility of a new or different kind 

of accident from any accident 

previously evaluated? 

Response: No. 

The proposed change in the 

surveillance test frequency does not 

create the possibility of a new or 

different kind of accident, since there 

are no physical changes being made to 

the plant and there are no changes to the 

operation of the plant that could 

introduce a new failure mode, creating 

an accident or affecting the mitigation of 

an accident. 

Therefore, the proposed change does 

not create the possibility of a new or 

different kind of accident from any 

accident previously analyzed. 

3. Does the proposed change involve 

a significant reduction in a margin of 

safety? 

Response: No. 

The proposed change revises the 

surveillance requirement to verify that 

the drywell spray headers and nozzles 

are unobstructed. Industry experience, 

Vermont Yankee surveillance history 

and the environmental conditions the 

system is subjected to are adequate to 

ensure continued system availability. As 

the spray nozzles are expected to remain 

unobstructed and be able to perform 

their post-accident function, plant safety 

is not affected. 

Therefore, the proposed change does 

not involve a significant reduction in a 

margin of safety. 

Based on this review, it appears that 

the three standards of 10 CFR 50.92 are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. David R. 

Lewis, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037–1128. 

NRC Section Chief: Allen G. Howe. 

Pacific Gas and Electric Company, 

Docket No. 50–133, Humboldt Bay 

Power Plant, Unit 3, Humboldt County, 

California 

Date of amendment request: July 9, 

2004. 

Description of amendment request:

The Humboldt Bay Power Plant (HBPP), 

Unit 3, is a decommissioning nuclear 

power plant that was permanently 

shutdown in July 1976. In December of 

2003, Pacific Gas and Electric (PG&E or 

the licensee) applied for a license to 

store its spent fuel in an onsite dry cask 

independent spent fuel storage 

installation (ISFSI). Moving the spent 

fuel to an ISFSI would permit the 

licensee to begin significant 

decommissioning activities. The 

licensee has chosen to use a Holtec HI–

STAR HB spent fuel cask handling 

system involving a spent fuel 

multipurpose canister and overpack. To 

facilitate spent fuel transfer from the 

HBPP spent fuel pool to the ISFSI, the 

licensee will also need to install a new 

crane that can be used to lift the cask 

handling system loaded with spent fuel 

assemblies. The licensee states it will be 

able to satisfy the applicable guidance of 

NUREG–0612, ‘‘Control of Heavy Loads 

at Nuclear Power Plants,’’ and NUREG–

0554. ‘‘Single-Failure Proof Cranes for 

Nuclear Power Plants,’’ in performing 

the necessary movement of the HBPP 

spent fuel to dry cask storage. The 

licensee has requested a license 

amendment that approves the use of the 

crane and associated changes to the 

HBPP Defueled Safety Analysis Report 

(DSAR) along with analyses, design, and 

procedural changes required to 

implement transfer of the spent fuel 

from the spent fuel pool to the ISFSI. 

Basis for proposed no significant 

hazards consideration determination:

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below:

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. With the HI–STAR HB System and the 

associated design and handling procedures, 

all cask drops and other events, which could 

damage other spent fuel, have been 

precluded through the robust handling 

systems, and mechanical arrangement that 

preclude crane movement over spent fuel, 

meeting the guidelines of NUREG–0612. 

Revisions of the HBPP procedures 

implementing the control of heavy loads 

ensures that PG&E will meet the NUREG–

0612 guidelines and will protect the fuel 

storage locations and the new HI–STAR HB 

System loading/unloading activities. As a 

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

offsite radiological release is not considered 

credible as demonstrated by the probabilistic 

evaluation that was performed using the 

guidelines of NUREG–0612 Appendix B and 

updated information from NUREG–1774 [‘‘A 

Survey of Crane Operating Experience at U.S. 

Nuclear Power Plants from 1968 through 

2002.’’] 

Other HBPP licensing-basis events, such as 

the drop of a spent fuel assembly, have not 

been affected by these changes and remain 

bounding events for potential radiological 

consequences. 

The proposed design of the dry cask 

system, the handling system, and associated 

procedural controls provide assurance that: 

(1) operational errors and mishandling 

events, and (2) support system malfunctions 

will not result in an increase in the 

probability or consequence of an accident 

previously analyzed. 

The proposed changes to use the Holtec 

HI–STAR HB system have been evaluated for 

seismic events and tornado missile impacts 

and it has been determined that these 

changes will not result in an increase in the 

probability or consequences of an accident 

previously evaluated. The Fire Protection 

Program will ensure that the combustible 

materials are properly controlled such that 

the total combustibles meet the current 

program commitments. Therefore, the 

proposed changes do not involve a 

76494 Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 

significant increase in the probability or 

consequences of an accident. 

2. Does the proposed amendment create 

the possibility of a new or different type of 

accident from any accident previously 

evaluated? 

No. The engineering design measures and 

the handling procedures preclude the 

possibility of new or different kinds of 

accidents. Damage to 10 CFR 50 structures, 

systems, and components from the cask 

handling and associated activities, and 

events resulting from possible damage to 

contained fuel have been considered. Both 

the types of accidents and the results remain 

within the envelope of existing HBPP DSAR 

licensing basis analyses, as demonstrated by 

the PG&E and Holtec analyses. 

The rupture of multipurpose canister 

(MPC) dewatering, forced helium 

dehydration or related closure system lines 

or the malfunction of equipment during cask 

handling operations resulting in radiological 

consequences are bounded by the HBPP 

DSAR fuel-handling accident analysis. 

Other design considerations, such as spent 

fuel pool (SFP) thermal, water chemistry and 

clarity, criticality, and structural, were 

evaluated and determined not to introduce 

the possibility of a new or different kind of 

accident from any previously evaluated. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

No. With the Holtec HI–STAR HB System, 

and the associated design and handling 

procedures, cask drops and other events have 

been precluded through robust load handling 

systems, providing defense-in-depth as 

described in NUREG–0612. Cask tipovers, 

while not considered credible, are shown to 

be below the 60g limit, preventing damage to 

the contained fuel assemblies (and associated 

structures), and meeting the analysis 

guidelines of NUREG–0612. As the existing 

licensing basis assumes a nonmechanistic 

drop damaging the SFP and all fuel, the 

result of this design approach with the 

minimization of drops and the associated 

structural challenges assure the margin of 

safety has been maintained. 

Other HBPP licensing-basis events, such as 

the drop of a spent fuel assembly, have not 

been affected by these changes and remain 

bounding events. Revision of HBPP 

procedures implementing the control of 

heavy loads to incorporate the additional 

restrictions on heavy loads movement will 

not affect the procedures or methodology 

used and will, therefore, not affect margins.

Adverse effects from seismic events and/or 

cask drops or tipovers have been evaluated, 

assuring that the fuel, MPC, and overpack 

remain within their design bases. Since 

design basis criteria are fully satisfied, there 

is no impact on the margin of safety. 

The Fire Protection Program will continue 

to ensure that the combustible materials are 

properly controlled such that the total 

combustibles meet the current program 

commitments. Thus, there are no significant 

reductions in margin of safety associated 

with these changes. 

Other design considerations, such as SFP 

thermal, water chemistry, criticality, and 

structural, were evaluated and determined to 

not involve a reduction in a margin of safety.

Based on the above evaluations, the 

licensee concludes that the activities 

associated with the above changes 

present no significant hazards 

consideration under the standards set 

forth in 10 CFR 50.92 and accordingly, 

a finding by the NRC of no significant 

hazards consideration is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Richard F. 

Locke, Esquire, Pacific Gas and Electric 

Company, P.O. Box 7442, San 

Francisco, California 94120. 

NRC Section Chief: Claudia Craig. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request: October 

6, 2004 

Brief description of amendments: The 

proposed change will revise the 

Technical Specification (TS) 3.8.1, ‘‘AC 

Sources—Operating,’’ to allow 

surveillance testing of the onsite diesel 

generators (DGs) during power 

operation. 

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 licensee’s analysis 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 design of plant equipment is not being 

modified by the proposed changes. In 

addition, the DGs and their associated 

emergency loads are accident mitigating 

features. As such, testing of the diesel 

generators (DGs) themselves is not associated 

with any potential accident-initiating 

mechanism. Therefore, there will be no 

significant impact on any accident 

probabilities by the approval of the requested 

changes. 

The changes include an increase in the 

online time that a DG under test will be 

paralleled to the grid (for SRs [Surveillance 

Requirements] 3.8.1.10 and 3.8.1.14) or 

unavailable due to testing (per SR 3.8.1.13). 

However, the overall time that the DG is 

paralleled in all modes (outage/non-outage) 

should remain unchanged. As such, the 

ability of the tested DG to respond to a design 

basis accident [DBA] could be adversely 

impacted by the proposed changes. However, 

the impacts are not considered significant 

based, in part, on the ability of the remaining 

DG to mitigate a DBA or provide safe 

shutdown. With regard to SR 3.8.1.10 and SR 

3.8.1.14, experience shows that testing per 

these SRs typically does not perturb the 

electrical distribution system and share the 

same electrical configuration alignment as 

the current monthly surveillance. In 

addition, operating experience and 

qualitative evaluation of the probability of 

the DG or bus loads being adversely affected 

concurrent with or due to a significant grid 

disturbance, while the DG is being tested, 

support the conclusion that the proposed 

changes do not involve any significant 

increase in the likelihood of a safety-related 

bus blackout or damage to plant loads. 

The SR changes that are consistent with 

TSTF [Technical Specification Task Force]-

283 have been approved generically and for 

individual Licensees. The on-line tests 

allowed by the TSTF are only to be 

performed for the purpose of establishing 

OPERABILITY. Performance of these SRs 

during restricted MODES will require an 

assessment to assure plant safety is 

maintained or enhanced. 

Deletion of expired TS LCO [Limiting 

Condition for Operation] 3.8.1, Required 

Action A.3, one-time 21-day Completion 

Time allowance for Startup Transformer 

XST2 preventive maintenance is an 

administrative change only. 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 accident 

from any accident previously evaluated? 

Response: No. 

The proposed changes would not create 

any new accidents since no changes are being 

made to the plant that would introduce any 

new accident causal mechanisms. Equipment 

will be operated in the same configuration as 

currently allowed for other DG SRs that allow 

testing during at-power operation. Deletion of 

expired TS LCO 3.8.1, Required Action A.3, 

one-time 21-day Completion Time allowance 

for Startup Transformer XST2 preventive 

maintenance is an administrative change 

only. This license amendment request does 

not impact any plant systems that are 

accident initiators; neither does it adversely 

impact any accident mitigating systems. 

Therefore, the proposed change does not 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not involve a 

significant reduction in the margin of safety. 

The margin of safety is related to the 

confidence in the ability of the fission 

product barriers to perform their design 

functions during and following an accident 

situation. These barriers include the fuel 

cladding, the reactor coolant system, and the 

containment system. The proposed changes 

do not directly affect these barriers, nor do 

Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 76495

they involve any significant adverse impact 

on the DGs which serve to support these 

barriers in the event of an accident 

concurrent with a loss of offsite power. The 

proposed changes to the testing requirements 

for the plant DGs do not affect the 

OPERABILITY requirements for the DGs, as 

verification of such OPERABILITY will 

continue to be performed as required (except 

during different allowed MODES). The 

changes have an insignificant impact on DG 

availability, as continued verification of 

OPERABILITY supports the capability of the 

DGs to perform their required function of 

providing emergency power to plant 

equipment that supports or constitutes the 

fission product barriers. Only one DG is to be 

tested at a time, so that the remaining DG 

will be available to safely shut down the 

plant if required. Consequently, performance 

of the fission product barriers will not be 

impacted by implementation of the proposed 

amendment. 

In addition, the proposed changes involve 

no changes to setpoints or limits established 

or assumed by the accident analysis. On this 

and the above basis, no safety margins will 

be impacted.

Deletion of expired TS LCO 3.8.1, Required 

Action A.3, one-time 21-day Completion 

Time allowance for Startup Transformer 

XST2 preventive maintenance is an 

administrative change only. 

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: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Section Chief: Mike Webb, 

Acting Chief. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request: October 

13, 2004. 

Brief description of amendments: The 

proposed changes will revise the 

Technical Specifications (TSs) to 

incorporate two topical reports used to 

determine the core operating limits of 

Comanche Peak Steam Electric Station 

(CPSES), Units 1 and 2, and delete 

reference to four topical reports and a 

reference to NUREG–0800 that are no 

longer required to support CPSES, Units 

1 and 2, core operating limits. 

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 licensee’s analysis 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 change is administrative in 

nature and as such does not impact the 

condition or performance of any plant 

structure, system or component. The core 

operating limits are established to support 

Technical Specifications 3.1, 3.2, 3.3, and 

3.4. The core operating limits ensure that fuel 

design limits are not exceeded during any 

conditions of normal operation or in the 

event of any Anticipated Operational 

Occurrence (AOO). The methods used to 

determine the core operating limits for each 

operating cycle are based on methods 

previously found acceptable by the NRC and 

listed in TS section 5.6.5.b. Application of 

these approved methods will continue to 

ensure that acceptable operating limits are 

established to protect the fuel cladding 

integrity during normal operation and AOOs. 

The requested Technical Specification 

changes do not involve any plant 

modifications or operational changes that 

could affect system reliability, performance, 

or possibility of operator error. The requested 

changes do not affect any postulated accident 

precursors, do not affect any accident 

mitigation systems, and do not introduce any 

new accident initiation mechanisms. 

As a result, the proposed change to the 

CPSES Technical Specifications does not 

involve any increase in the probability or the 

consequences of any accident or malfunction 

of equipment important to safety previously 

evaluated since neither accident probabilities 

nor consequences are being affected by this 

proposed administrative change. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change is administrative in 

nature, and therefore does not involve any 

changes in station operation or physical 

modifications to the plant. In addition, no 

changes are being made in the methods used 

to respond to plant transients that have been 

previously analyzed. No changes are being 

made to plant parameters within which the 

plant is normally operated or in the 

setpoints, which initiate protective or 

mitigative actions, and no new failure modes 

are being introduced. 

Therefore, the proposed administrative 

change to the CPSES Technical 

Specifications does not create the possibility 

of a new or different kind of accident or 

malfunction of equipment important to safety 

from any previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change is administrative in 

nature and does not impact station operation 

or any plant structure, system or component 

that is relied upon for accident mitigation. 

Furthermore, the margin of safety assumed in 

the plant safety analysis is not affected in any 

way by the proposed administrative change. 

Therefore, the proposed change to the 

CPSES Technical Specifications does not 

involve any 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: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Section Chief: Michael Webb, 

Acting Chief. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice.

Florida Power and Light Company, et 

al., Docket No. 50–389, St. Lucie Plant, 

Unit No. 2, St. Lucie County, Florida 

Date of amendment request:

November 8, 2004. 

Description of amendment request: To 

revise Technical Specification Section 

4.4.5.4 to modify the definitions of 

steam generator (SG) tube ‘‘Plugging 

Limit’’ and ‘‘Tube Inspection.’’

Date of publication of individual 

notice in the Federal Register:

November 24, 2004 (69 FR 68408). 

Expiration date of individual notice:

January 24, 2005. 

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. 

76496 Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

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 PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov.

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of application for amendment:

April 1, 2004. 

Brief description of amendment: The 

amendment revises Technical 

Specification (TS) requirements to adopt 

the provisions of the TS Task Force 

(TSTF) change TSTF–359, regarding 

increased flexibility in mode changes. 

The availability of TSTF–359 for 

adoption by licensees was announced in 

the Federal Register on April 4, 2003 

(68 FR 16579). 

Date of issuance: November 29, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment No.: 163. 

Facility Operating License No. NPF–

43: Amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 24, 2004 (69 FR 

52037).

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated November 29, 

2004. 

No significant hazards consideration 

comments received: No. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of amendment request: October 

7, 2004, as supplemented by letters 

dated November 12 and 18, 2004. 

Description of amendment request:

The amendment revised the Safety 

Limit Minimum Critical Power Ratio in 

Technical Specification 2.1.1.2 to reflect 

the results of cycle-specific calculations 

performed for Fermi 2 operating Cycles 

10 and 11. 

Date of issuance: November 30, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to startup for Fermi 2 Cycle 11 

operation. 

Amendment No.: 164. 

Facility Operating License No. NPF–

43: Amendment revises the Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): Yes. (November 

9, 2005; 69 FR 64986) The notice 

provided an opportunity to submit 

comments on the Commission’s 

proposed NSHC determination. No 

comments have been received. The 

notice also provided an opportunity to 

request a hearing by January 10, 2005, 

but indicated that if the Commission 

makes a final NSHC determination, any 

such hearing would take place after 

issuance of the amendment. 

The Commission’s related evaluation 

of the amendment, finding of exigent 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated November 

30, 2004. 

Attorney for licensee: Peter 

Marquardt, Legal Department, 688 WCB, 

Detroit Edison Company, 2000 2nd 

Avenue, Detroit, Michigan 48226–1279. 

NRC Section Chief: L. Raghavan. 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

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

2), Beaver County, Pennsylvania 

Date of application for amendments:

January 26, 2004, as supplemented 

September 13, 2004. 

Brief description of amendments:

These amendments authorized changes 

to the BVPS–1 and 2 Updated Final 

Safety Analysis Reports (UFSARs) to 

revise the level of the Ohio River that is 

assumed at the onset of an accident 

during power operation to be 654.0′

mean sea level (msl) instead of 649.0′

msl for BVPS–1 and 2. The proposed 

change is consistent with current 

Technical Specification 3.7.5.1, which 

requires the plant to shut down when 

the Ohio River reaches a level below 

654.0′ msl. 

Date of issuance: November 29, 2004. 

Effective date: As of the date of 

issuance and shall submit the changes 

authorized by these amendments with 

the next update of the UFSARs in 

accordance with 10 CFR 50.71(e). 

Amendment Nos.: 264 and 145. 

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments authorize 

changes to the UFSARs. 

Date of initial notice in Federal 

Register: March 16, 2004 (69 FR 

12369). 

The supplement dated September 13, 

2004, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 29, 

2004 . 

No significant hazards consideration 

comments received: No. 

Florida Power and Light Company, et 

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

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

County, Florida 

Date of application for amendments:

November 21, 2003, as supplemented by 

letters dated May 18, and August 23, 

2004. 

Brief description of amendments:

These amendments revised the St. Lucie 

Unit 1 and 2 Technical Specifications 

(TSs) to eliminate certain pressure 

sensor response time testing 

requirements. Elimination of these tests 

is discussed in the Combustion 

Engineering Owners Group Topical 

Report CE NPSD–1167, Revision 2, 

‘‘Elimination of Pressure Sensor 

Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 76497

Response Time Testing Requirements,’’

which was approved by the NRC staff in 

letters dated July 24, 2000, and 

December 5, 2000. Specifically, these 

amendments revise the St. Lucie Units 

1 and 2 TS Definitions 1.12, 

‘‘Engineered Safety Features Response 

Time,’’ and 1.26, ‘‘Reactor Protection 

System Response Time.’’

Date of Issuance: November 30, 2004. 

Effective Date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: 195 and 137 

Renewed Facility Operating License 

Nos. DPR–67 and NPF–16: Amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: September 28, 2004 (69 FR 

57675). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 30, 

2004. 

No significant hazards consideration 

comments received: No.

Florida Power and Light Company, 

Docket Nos. 50–250 and 50–251, Turkey 

Point Plant, Units 3 and 4, Miami-Dade 

County, Florida 

Date of application for amendments:

November 26, 2002, as supplemented by 

letters dated September 8, 2003, October 

30, 2003, June 21, 2004, and October 8, 

2004. 

Brief description of amendments: The 

amendments increased the total spent 

fuel wet storage capacity for each unit, 

by adding a spent fuel storage rack in 

the cask area in each unit’s spent fuel 

pool. Each rack increased both units’

storage capacity by 131 fuel assemblies. 

The amendments also included the 

addition of the design of the racks in 

Section 5.6.1.1.c of the Technical 

Specifications (TSs), and revised the 

stated spent fuel capacity in TS Section 

5.6.3 and the location called out in the 

Design Features Sections 5.6.1.1a and b 

of the TSs referring to Updated Final 

Safety Analysis Report Appendix 14D 

rather the Westinghouse Report WCAP–

14416–P. 

Date of issuance: November 24, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos: 226 and 222. 

Renewed Facility Operating License 

Nos. DPR–31 and DPR–41: Amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: January 28, 2003 (69 FR 

4246). 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 amendments is contained in an 

Environmental Assessment dated 

October 17, 2003, and in a Safety 

Evaluation dated November 24, 2004. 

No significant hazards consideration 

comments received: No. 

Nine Mile Point Nuclear Station, LLC, 

Docket No. 50–220, Nine Mile Point 

Nuclear Station, Unit No. 1, Oswego 

County, New York 

Date of application for amendment:

August 17, 2004. 

Brief description of amendment: The 

amendment revised Section 3.3.1, 

‘‘Oxygen Concentration,’’ of the 

Technical Specifications to add a new 

action, allowing 24 hours to restore the 

oxygen concentration within the limit of 

<4% by volume if the limit is exceeded 

when the reactor is operating in the 

power operating condition. 

Date of Issuance: November 29, 2004. 

Effective date: November 29, 2004 and 

shall be implemented within 15 days of 

issuance. 

Amendment No.: 185. 

Facility Operating License No. DPR–

63: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 31, 2004 (69 FR 

53110). 

The Commission’s related evaluation 

of this amendment is contained in a 

Safety Evaluation dated November 29, 

2004. 

No significant hazards consideration 

comments received: No. 

Pacific Gas and Electric Company, 

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

Canyon Nuclear Power Plant, Unit Nos. 

1 and 2, San Luis Obispo County, 

California 

Date of application for amendments:

September 12, 2003, and its 

supplements dated April 23, June 4, and 

August 30, 2004. 

Brief description of amendments: The 

amendments increase the current steam 

generator narrow range water level-low 

low setpoints from greater or equal to 

7.0 percent allowable value and 7.2 

percent nominal trip setpoint to greater 

than or equal to 14.8 percent allowable 

value and 15.0 percent nominal trip 

setpoint. The reactor trip setpoint is 

specified in TS Table 3.3.1–1, ‘‘Reactor 

Trip System Instrumentation,’’ and the 

actuation setpoint to start the auxiliary 

feedwater pumps is specified in TS 

Table 3.3.2–1, ‘‘Engineered Safety 

Feature Actuation System 

Instrumentation.’’

Date of issuance: December 2, 2004. 

Effective date: December 2, 2004, and 

shall be implemented within 90 days 

from the date of issuance. 

Amendment Nos.: Unit 1–178; Unit 

2–180. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: November 25, 2003 (68 FR 

66138) The April 23, June 4, and August 

30, 2004, supplemental letters provided 

additional clarifying information, did 

not expand the scope of the application 

as originally noticed, and did not 

change the staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated December 2, 

2004. 

No significant hazards consideration 

comments received: No. 

PPL Susquehanna, LLC, Docket Nos. 50–

387 and 50–388, Susquehanna Steam 

Electric Station, Units 1 and 2 (SSES 1 

and 2), Luzerne County, Pennsylvania 

Date of application for amendments:

December 5, 2003, as supplemented by 

letter dated June 4, 2004. 

Brief description of amendments: The 

amendments revised SSES 1 and 2 

Technical Specifications (TSs) by 

adding a requirement to apply linear 

heat generation (LHGR) limits if the 

main turbine bypass system becomes 

inoperable. The proposed changes 

clarify TS 3.7.6 to state that both 

minimum critical power ratio and LHGR 

limits for an inoperable main turbine 

bypass system are required if the system 

becomes inoperable.

Date of issuance: December 3, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 218 and 193. 

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: January 6, 2004 (69 FR 698). 

The supplement dated June 6, 2004, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated December 3, 

2004. 

No significant hazards consideration 

comments received: No. 

76498 Federal Register / Vol. 69, No. 244 / Tuesday, December 21, 2004 / Notices 

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 

12, 2004, as superseded by letter dated 

October 5, 2004, as supplemented by 

letter dated October 11, 2004. 

Brief description of amendments: The 

amendments revise Technical 

Specification (TS) 3/4.4.5, in 

conjunction with the new 

administrative control TS 6.8.3.o and 

reporting requirement TS 6.9.1.7, to 

establish a new programmatic, largely 

performance-based framework for 

ensuring SG tube integrity. The reactor 

coolant system leakage requirements of 

TS 3.4.6.2 are also revised. 

Date of issuance: November 24, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: Unit 1—164; Unit 

2—154. 

Facility Operating License Nos. NPF–

76 and NPF–80: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: August 31, 2004 (69 FR 

53113). The October 5, 2004, letter 

which superseded the August 12, 2004, 

letter and the supplement dated October 

11, 2004, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not significantly change the 

staff’s original proposed no significant 

hazards consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 24, 

2004. 

No significant hazards consideration 

comments received: No. 

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. 

Brief description of amendments: The 

amendments eliminate the requirements 

in the TS associated with hydrogen 

recombiners and hydrogen 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). 

Date of issuance: November 30, 2004. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: Unit 1—165; Unit 

2—155. 

Facility Operating License Nos. NPF–

76 and NPF–80: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: September 28, 2004 (69 FR 

57996). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 30, 

2004. 

No significant hazards consideration 

comments received: No.

Dated at Rockville, Maryland, this 13th day 

of December 2004.

For the Nuclear Regulatory Commission. 

James E. Lyons, 

Deputy Director, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation.

[FR Doc. 04–27614 Filed 12–20–04; 8:45 am] 

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE 

COMMISSION 

[Release Nos. 33–8514; 34–50864; File No. 

265–23] 

Advisory Committee on Smaller Public 

Companies

AGENCY: Securities and Exchange 

Commission.

ACTION: Notice of establishment of the 

Advisory Committee on Smaller Public 

Companies. 

SUMMARY: The Chairman of the 

Securities and Exchange Commission 

(‘‘Commission’’), with the concurrence 

of the other Commissioners, intends to 

establish the Securities and Exchange 

Commission Advisory Committee on 

Smaller Public Companies to assist the 

Commission in evaluating the current 

securities regulatory system relating to 

disclosure, financial reporting, internal 

controls, and offering exemptions for 

smaller public companies.

FOR FURTHER INFORMATION CONTACT:

Gerald J. Laporte, Chief, or Kevin M. 

O’Neill, Special Counsel, at (202) 942–

2950, Office of Small Business Policy, 

Division of Corporation Finance, 

Securities and Exchange Commission, 

450 Fifth Street, NW., Washington, DC 

20549–0310.

SUPPLEMENTARY INFORMATION: In 

accordance with the requirements of the 

Federal Advisory Committee Act, 5 

U.S.C. App. 1, the Securities and 

Exchange Commission (‘‘Commission’’) 

is publishing this notice that the 

Chairman of the Commission, with the 

concurrence of the other 

Commissioners, intends to establish the 

Securities and Exchange Commission 

Advisory Committee on Smaller Public 

Companies (the ‘‘Committee’’). The 

Committee’s objective is to assess the 

impact of the current regulatory system 

for smaller companies under the 

securities laws of the United States and 

make recommendations for changes. 

To achieve the Committee’s goals, 

between 11 and 21 members will be 

appointed who can represent effectively 

the varied interests affected by the range 

of issues to be considered. The 

Committee’s membership may include 

officers and directors of smaller 

companies; accountants, lawyers and 

other professional service providers to 

smaller companies; regulators; 

investors; and members of the public at 

large. The Committee’s membership will 

be fairly balanced in terms of the points 

of view represented and the functions to 

be performed. 

The Committee may be established 15 

days after publication of this notice in 

the Federal Register by filing a charter 

for the Committee complying with the 

Federal Advisory Committee Act with 

the Committee on Banking, Housing, 

and Urban Affairs of the United States 

Senate and the Committee on Financial 

Services of the United States House of 

Representatives. A copy of the charter 

also will be filed with the Chairman of 

the Commission, furnished to the 

Library of Congress, placed in the Public 

Reference Room at the Commission’s 

headquarters and posted on the 

Commission’s Internet Web site at 

www.sec.gov/info/smallbus.shtml. The 

Committee’s charter is expected to 

direct it to consider the following areas, 

including the impact in each area of the 

Sarbanes-Oxley Act of 2002, Pub. L. 

107–204, 116 Stat. 745 (July 30, 2002): 

• Corporate disclosure and reporting 

requirements and federally-imposed 

corporate governance requirements for 

smaller public companies, including 

differing regulatory requirements based 

on market capitalization, other 

measurements of size or market 

characteristics;

• Accounting standards and financial 

reporting requirements applicable to 

smaller public companies; 

• Frameworks for internal control over 

financial reporting applicable to smaller 

public companies, methods for 

management’s assessment of such 

internal control, and standards for 

auditing such internal control; and 

•The process, requirements and 

exemptions relating to offerings of 

securities by smaller companies, 

particularly public offerings. 

The charter will direct the Committee 

to conduct its work with a view to 

protecting investors, considering 

whether the costs imposed by the 

current securities regulatory system for