Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 46929 

Program Directors (CRCPD) (Public 

Meeting) (Contact: Shawn Smith, 

301–415–2620). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov. 

1 p.m. Discussion of Security Issues 

(Closed—Ex. 1). 

Week of September 18, 2006—Tentative 

There are no meetings scheduled for 

the Week of September 18, 2006. 

* The schedule for Commission 

meetings is subject to change on short 

notice. To verify the status of meetings 

call (recording)—(301) 415–1292. 

Contact person for more information: 

Michelle Schroll, (301) 415–1662. 

The NRC Commission Meeting 

Schedule can be found on the Internet 

at: www.nrc.gov/what-we-do/policymaking/schedule.html. 

The NRC provides reasonable 

accommodation to individuals with 

disabiloities where appropriate. If you 

need a reasonable acommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g., 

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

Deborah Chan, at 301–415–7041, TDD: 

301–415–2100, or by E-mail at 

DLC@nrc.gov. Determinations on 

requests for reasonable accommodation 

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

This notice is distributed by mail to 

several hundred subscribers; if you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301–415–1969). 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to dkw@nrc.gov. 

Dated: August 10, 2008. 

R. Michelle Schroll, 

Office of the Secretary. 

[FR Doc. 06–6939 Filed 8–11–06; 9:59 am] 

BILLING CODE 7590–01–M 

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from July 21, 

2006, to August 3, 2006. The last 

biweekly notice was published on 

August 1, 2006 (71 FR 43528). 

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 

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46930 Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

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. 

AmerGen Energy Company, LLC, Docket 

No. 50–461, Clinton Power Station 

(CPS), Unit 1, DeWitt County, Illinois 

Date of amendment request: June 30, 

2006. 

Description of amendment request: 

The proposed change would revise the 

Note preceding Technical Specification 

(TS) Surveillance Requirement (SR) 

3.4.6.1 to be consistent with the 

wording in NUREG–1434, ‘‘Standard 

Technical Specifications General 

Electric Plants, BWR/6,’’ Revision 3. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment revises the note 

associated with TS SR 3.4.6.1, which requires 

verification that the leakage past the Reactor 

Coolant System (RCS) Pressure Isolation 

Valves (PIVs) is less than a specified limit. 

The proposed revision provides clarification 

that performance of this SR is allowed during 

plant shutdown (i.e., a Mode other than 

Modes 1 and 2). 

The proposed change does not require 

modification to the facility. The proposed 

change does not affect the operation of any 

facility equipment, the interface between 

facility systems, or the reliability of any 

equipment. In addition, the proposed change 

does not alter the requirement to perform the 

leakage testing of the RCS PIVs and does not 

revise the leakage limits associated with this 

SR. The function of the RCS PIVs is to 

separate the high pressure RCS from an 

attached low pressure system. Periodic 

testing of PIVs can substantially reduce 

intersystem Loss of Coolant Accident (LOCA) 

probability. Since the proposed change does 

not alter the method or limits associated with 

the leak rate testing of the RCS PIVs there is 

no significant increase in the probability of 

a LOCA. Therefore, the proposed amendment 

does not involve a significant increase in the 

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Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 46931 

probability of an accident previously 

evaluated. 

The consequences of a previously analyzed 

event are dependent on the initial conditions 

assumed in the analysis, the availability and 

successful functioning of equipment assumed 

to operate in response to the analyzed event, 

and the setpoints at which these actions are 

initiated. The method for performing the 

leakage testing of the RCS PIVs and the 

specified leakage limit for this testing will 

not change as a result of the proposed 

revision and, therefore, there is no change in 

the consequences associated with the LOCA 

analysis. The radiological consequences 

remain within applicable regulatory limits. 

The proposed change does not alter any 

system’s performance measures or the ability 

to perform its accident mitigation functions. 

The radiological consequences associated 

with any previously evaluated accident do 

not change as a result of the proposed 

revision. Therefore, the proposed change 

does not involve a significant increase in the 

consequences of an accident previously 

evaluated. 

Based on the above, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change to the wording of the 

Note to TS SR 3.4.6.1 clarifies the plant 

conditions for when the surveillance is 

required to be performed. The proposed 

change does not affect the design, functional 

performance or operation of the facility. No 

new equipment is being introduced and 

installed equipment is not being operated in 

a new or different manner. Similarly, the 

proposed change does not affect the design 

or operation of any structures, systems or 

components involved in the mitigation of any 

accidents, nor does it affect the design or 

operation of any component in the facility 

such that new equipment failure modes are 

created. There are no setpoints at which 

protective or mitigative actions are initiated 

that are affected by this proposed action. No 

change is being made to procedures relied 

upon to respond to an off-normal event. 

As such the proposed amendment will not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Margins of safety are established in the 

design of components, the configuration of 

components to meet certain performance 

parameters, and in the establishment of 

setpoints to initiate alarms or actions. The 

proposed change revises a note associated 

with a surveillance requirement to clarify the 

plant conditions for when the surveillance 

needs to be performed. This change involves 

an administrative clarification to reflect the 

original intent of the TS. The equipment will 

continue to be tested in a manner and at a 

frequency necessary to provide confidence 

that the equipment can perform its intended 

safety function. There is no change in the 

design of the affected systems, no alteration 

of the setpoints at which alarms or actions 

are initiated, and no change in plant 

configuration from original design. There is 

no impact on the plant safety analyses. 

Therefore, operation of CPS in accordance 

with the proposed change will 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. Bradley J. 

Fewell, Assistant General Counsel, 

Exelon Generation Company, LLC, 200 

Exelon Way, Kennett Square, PA 19348. 

NRC Branch Chief: Daniel S. Collins. 

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: June 14, 

2006. 

Description of amendment request: 

The proposed change will delete 

Waterford 3 Technical Specification 

(TS) Surveillance Requirement (SR) 

4.8.1.1.2.f. This SR requires that the 

emergency diesel generator be subjected 

to an inspection in accordance with 

procedures prepared in conjunction 

with its manufacturer’s 

recommendations. 

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 ability of the emergency diesel 

generator to perform its safety function is not 

proven by the performance of the 

manufacturer’s recommended inspections. 

The inspections are not considered an 

initiator or mitigating factor in any 

previously evaluated accidents. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change results in the 

deletion of the SR associated with the 

performance of manufacturer’s inspections. 

No modifications to plant structures, 

systems, or components, or changes in the 

design of the plant structures, systems, or 

components are required to support the 

proposed TS change. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The ability of the emergency diesel 

generator to perform its safety function is not 

proven by the performance of the 

manufacturer’s recommended inspections. 

Inspection activities will continue to be 

performed. 

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: N.S. Reynolds, 

Esquire, Winston & Strawn, 1700 K 

Street, NW., Washington, DC 20006–

3817. 

NRC Branch Chief: David Terao. 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station (DNPS), 

Units 2 and 3, Grundy County, Illinois 

Date of amendment request: June 2, 

2006. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specification (TS) 

Surveillance Requirement (SR) 3.4.3.1 to 

increase the allowable as-found main 

steam safety valve (MSSV) lift set point 

tolerance from +/¥1 percent to +/¥3 

percent. The proposed change would 

also revise the SR 3.1.7.10 to increase 

the enrichment of sodium pentaborate 

used in the Standby Liquid Control 

(SLC) system from greater than or equal 

to 30 atom percent boron-10 to greater 

than or equal to 45 atom percent boron10. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change increases the 

allowable as-found MSSV lift setpoint 

tolerance, determined by test after the valves 

have been removed from service, from +/¥1 

percent to +/¥3 percent. The proposed 

change does not alter the TS requirements for 

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46932 Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 

the number of MSSVs required to be 

operable, the nominal lift setpoints, the 

allowable as-left lift setpoint tolerance, the 

MSSV testing frequency, or the manner in 

which the valves are operated. 

Consistent with current TS requirements, 

the proposed change continues to require 

that the MSSVs be adjusted to within +/¥1 

percent of their nominal lift setpoints 

following testing. Since the proposed change 

does not alter the manner in which the valves 

are operated, there is no significant impact 

on reactor operation. 

The proposed change does not involve a 

physical change to the valves, nor does it 

change the safety function of the valves. The 

proposed TS revision involves no significant 

changes to the operation of any systems or 

components in normal or accident operating 

conditions and no changes to existing 

structures, systems, or components, with the 

exception of the SLC system enrichment 

change. The proposed change to increase the 

enrichment of sodium pentaborate used in 

the SLC system by a design modification 

using a single SLC pump will ensure that the 

requirements of 10 CFR 50.62, 

‘‘Requirements for reduction of risk from 

anticipated transients without scram (ATWS) 

events for light-water-cooled nuclear power 

plants,’’ continue to be met. The SLC system 

is not an initiator to an accident; rather, the 

SLC system is used to mitigate a postulated 

anticipated transient without scram (ATWS) 

event. Therefore, these changes will not 

increase the probability of an accident 

previously evaluated. 

Generic considerations related to the 

change in setpoint tolerance were addressed 

in NEDC–31753P, ‘‘BWROG In-Service 

Pressure Relief Technical Specification 

Revision Licensing Topical Report,’’ and 

were reviewed and approved by the NRC in 

a safety evaluation dated March 8, 1993. The 

plant specific evaluations, required by the 

NRC’s safety evaluation and performed to 

support this proposed change, show that 

there is no change to the design core thermal 

limits and adequate margin to the reactor 

vessel pressure limits using a +/¥3 percent 

lift setpoint tolerance. These analyses also 

show that operation of Emergency Core 

Cooling Systems is not affected, and the 

containment response following a loss-ofcoolant accident is acceptable. The plant 

systems associated with these proposed 

changes are capable of meeting applicable 

design basis requirements and retain the 

capability to mitigate the consequences of 

accidents described in the Updated Final 

Safety Analysis Report. Therefore, these 

changes do not involve an increase in the 

consequences of an accident previously 

evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change increases the 

allowable as-found lift setpoint tolerance for 

the DNPS MSSVs, and increases the required 

enrichment of sodium pentaborate used in 

the SLC system. The proposed change to 

increase the enrichment of sodium 

pentaborate used in the SLC system will 

ensure that the requirements of 10 CFR 50.62 

continue to be met. 

The proposed change to increase the MSSV 

tolerance was developed in accordance with 

the provisions contained in the NRC safety 

evaluation for NEDC–31753P. MSSVs 

installed in the plant following testing or 

refurbishment will continue to meet the 

current tolerance as-left acceptance criteria of 

+/¥1 percent of the nominal setpoint. The 

proposed change does not affect the manner 

in which the overpressure protection system 

is operated; therefore, there are no new 

failure mechanisms for the overpressure 

protection system. 

The proposed change to allow an increase 

in the MSSV setpoint tolerance does not alter 

the nominal MSSV lift setpoints or the 

number of MSSVs currently required to be 

operable by DNPS TS. The proposed change 

does not involve physical changes to the 

valves, nor does it change the safety function 

of the valves. There is no alteration to the 

parameters within which the plant is 

normally operated. As a result, no new 

failure modes are being introduced. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The margin of safety is established through 

the design of the plant structures, systems, 

and components, the parameters within 

which the plant is operated, and the 

establishment of the setpoints for the 

actuation of equipment relied upon to 

respond to an event. The proposed change 

does not modify the safety limits or setpoints 

at which protective actions are initiated, and 

does not change the requirements governing 

operation or availability of safety equipment 

assumed to operate to preserve the margin of 

safety. 

Establishment of the ±3 percent MSSV 

setpoint tolerance limit does not adversely 

impact the operation of any safety-related 

component or equipment. Evaluations 

performed in accordance with the NRC safety 

evaluation for NEDC–31753P have concluded 

that all design limits will continue to be met. 

The proposed change to increase the 

enrichment of sodium pentaborate used in 

the SLC system will ensure that the 

requirements of 10 CFR 50.62 continue to be 

met. 

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 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Assistant General Counsel, 

Exelon Generation Company, LLC, 200 

Exelon Way, Kennett Square, PA 19348. 

NRC Branch Chief: Daniel S. Collins. 

Exelon Generation Company, LLC, 

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

County Station (LSCS), Units 1 and 2, 

LaSalle County, Illinois 

Date of amendment request: March 

16, 2006. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification (TS) 

3.3.6.1, ‘‘Primary Containment Isolation 

Instrumentation,’’ Table 3.3.6.1–1 to 

revise the allowable values (AVs) for the 

reactor core isolation cooling (RCIC) 

temperature-based leak detection. The 

proposed change is a result of revising 

the setpoint calculation for the subject 

temperature instruments based on the 

current reactor coolant leak detection 

analytical limit. The temperature limits 

correspond to a 25-gallon per minute 

(gpm) leak as determined by LSCS 

calculations. The proposed changes 

would revise TS Table 3.3.6.1–1 AVs for 

the following four RCIC system isolation 

functions: 

Item 3.e. RCIC Equipment Room 

Temperature—High 

Item 3.f. RCIC Equipment Room Differential 

Temperature—High 

Item 3.g. RCIC Steam Line Tunnel 

Temperature—High 

Item 3.h. RCIC Steam Line Tunnel 

Differential Temperature—High 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The proposed change is a result of revising 

the setpoint calculation for the subject 

temperature instruments based on the current 

reactor coolant leak detection calculation 

analytical limit. The proposed changes will 

revise TS Table 3.3.6.1–1 Allowable Values 

for the following four RCIC system isolation 

functions as noted below. 

• Increase the Allowable Value for Function 

3.e., ‘‘RCIC Equipment Room 

Temperature—High,’’ from ≤ 291.0 °F to ≤ 

297.0 °F 

• Decrease the Allowable Value for Function 

3.f., ‘‘RCIC Equipment Room Differential 

Temperature—High,’’ from ≤ 189.0 °F to ≤ 

188.0 °F 

• Decrease the Allowable Value for Function 

3.g., ‘‘RCIC Steam Line Tunnel 

Temperature—High,’’ from ≤ 277.0 °F to ≤ 

267.0 °F 

• Increase the Allowable Value for Function 

3.h., ‘‘RCIC Steam Line Tunnel Differential 

Temperature—High,’’ from ≤ 155.0 °F to ≤ 

163.0 °F 

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Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 46933 

The function of the instrumentation listed 

on TS Table 3.3.6.1–1, in combination with 

other accident mitigation features, is to limit 

fission product release during and following 

postulated Design Basis Accidents to within 

allowable limits. The Allowable Values 

specified in TS Table 3.3.6.1–1 provide 

assurance that the instrumentation will 

perform as designed. 

The Allowable Values for RCIC system 

isolation are not a precursor to any accident 

previously evaluated. Accidents are assumed 

to be initiated by equipment failure. The 

proposed change does not alter the initiation 

conditions or operational parameters for the 

system. There is no increase in the failure 

probability of the system. As such, the 

probability of occurrence for a previously 

evaluated accident is not increased. 

The Allowable Values specified in Table 

3.3.6.1–1 provide assurance that the RCIC 

system will perform as designed. The 

proposed revision to the Allowable Values 

does not change any of the RCIC system leak 

detection isolation actuation setpoints. Thus, 

the radiological consequences of any 

accident previously evaluated are not 

increased. 

Based on the above information, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. The proposed TS change does not create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change does not affect the 

control parameters governing unit operation 

or the response of plant equipment to 

transient conditions. The proposed change 

does not change or introduce any new 

equipment, modes of system operation or 

failure mechanisms. 

The proposed change is based on revised 

reactor coolant leak detection calculation 

analytical limits determined by the most 

current revision to the heat rise calculation. 

Setpoint calculations have been performed to 

determine the nominal trip setpoints and 

Allowable Values for the instrumentation 

associated with the leak detection function 

based on the revised analytical limits 

determined by the heat rise calculations. The 

proposed revision to the Allowable Values 

does not change any of the RCIC system leak 

detection isolation actuation setpoints. 

Based on the above information, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. The proposed change does not involve 

a significant reduction in the margin of 

safety. 

The proposed change will revise TS Table 

3.3.6.1–1 Allowable Values for the 

instrument functions associated with RCIC 

Isolation. 

The current Allowable Values for these 

functions are: 

≤ 291.0 °F for RCIC Equipment Room 

Temperature—High 

≤ 189.0 °F for RCIC Equipment Room 

Differential Temperature—High 

≤ 277.0 °F for the RCIC Steam Line Tunnel 

Temperature—High 

≤ 155.0 °F for the RCIC Steam Line Tunnel 

Differential Temperature—High 

The proposed change revises the Allowable 

Values to the following: 

≤ 297.0 °F for RCIC Equipment Room 

Temperature—High 

≤ 188.0 °F for RCIC Equipment Room 

Differential Temperature—High 

≤ 267.0 °F for the RCIC Steam Line Tunnel 

Temperature—High 

≤ 163.0 °F for the RCIC Steam Line Tunnel 

Differential Temperature—High 

The proposed change is a result of revising 

the setpoint calculation for the subject 

temperature instruments based on the current 

analytical limit. The proposed changes will 

revise TS Table 3.3.6.1–1 Allowable Values 

for the subject four RCIC system isolation 

functions and will provide assurance that the 

RCIC system will perform as designed. The 

proposed revision to the Allowable Values 

does not change any of the RCIC system leak 

detection isolation actuation setpoints. 

Margin of safety is established by the 

design and qualification of plant equipment, 

the operation of the plant within analyzed 

limits, and the point at which protective or 

mitigative actions are being initiated. The 

proposed change does not alter these 

considerations. The proposed allowable 

values will still ensure that the results of the 

accident analysis remain valid. 

Based on this information, 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 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Assistant General Counsel, 

Exelon Generation Company, LLC, 200 

Exelon Way, Kennett Square, PA 19348. 

NRC Branch Chief: Daniel S. Collins. 

Exelon Generation Company, LLC, 

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

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Date of amendment request: April 4, 

2006. 

Description of amendment request: 

The proposed amendment request will 

add one NRC approved topical report 

reference to the list of analytical 

methods in Technical Specification (TS) 

5.6.5, ‘‘Core Operating Limits Report 

(COLR),’’ that can be used to determine 

core operating limits, and will delete 

seven obsolete references from the same 

TS Section. 

The proposed changes are: 

1. Add an NRC previously approved 

Topical Report ANF–1358(P)(A), Revision 3, 

‘‘The Loss of Feedwater Heating Transient in 

Boiling Water Reactors,’’ (LOFWH), which 

will list FRA–ANP method for evaluating the 

LOFWH transient. 

2. Delete seven references describing 

previously approved Global Nuclear Fuel 

(GNF) and FRA–ANP methodologies for the 

analyses of ATRIUM–9B and GE9 fuel. Both 

of these fuel types have been or will be 

completely discharged from both Lasalle 

County Station (LSCS) reactors after the 

loading of ATRIUM–10 fuel during the LSCS 

Unit 2 refuel outage currently scheduled to 

begin in February 2007 (i.e., L2R11). 

The proposed changes support the 

continued irradiation of ATRIUM–10 

fuel in the LSCS reactors and the use of 

the NRC-approved analytical 

methodology for evaluation of LOFWH 

transients. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

Technical Specification (TS) 5.6.5 lists 

NRC-approved analytical methods used at 

LaSalle County Station (LSCS) to determine 

core operating limits. The proposed changes 

will add an NRC-approved topical report 

reference to the list of administratively 

controlled analytical methods in TS 5.6.5, 

‘‘Core Operating Limits Report (COLR),’’ that 

can be used to determine core operating 

limits, and delete seven obsolete references. 

The addition of a Framatome ANP (FRA–

ANP) methodology to determine overall core 

operating limits for future LSCS core 

configurations was approved by the NRC in 

Reference 2. LSCS Unit 2 will continue to 

load Framatome ANP ATRIUM–10 fuel 

during the Unit 2 Refueling Outage 11 

currently scheduled for February 2007. The 

proposed change to TS 5.6.5 will add a FRA–

ANP methodology as a reference to 

determine core operating limits for loss of 

feedwater heater (LOFWH) conditions. Thus, 

the proposed change will allow LSCS to use 

the most recent FRA–ANP methodology for 

analysis of LOFWH conditions. 

The addition and deletion of approved 

analytical methods in TS Section 5.6.5 has no 

effect on any accident initiator or precursor 

previously evaluated and does not change the 

manner in which the core is operated. The 

NRC-approved methods ensure that the 

output accurately models predicted core 

behavior, have no effect on the type or 

amount of radiation released, and have no 

effect on predicted offsite doses in the event 

of an accident. Additionally, the NRCapproved methods do not change any key 

core parameters that influence any accident 

consequences. Thus, the proposed changes 

do not have any effect on the probability of 

an accident previously evaluated. 

The methodology conservatively 

establishes acceptable core operating limits 

such that the consequences of previously 

analyzed events are not significantly 

increased. 

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46934 Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 

The proposed changes in the list of 

analytical methods do not affect the ability of 

LSCS to successfully respond to previously 

evaluated accidents and does not affect 

radiological assumptions used in the 

evaluations. Thus, the radiological 

consequences of any accident previously 

evaluated are not increased. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes to TS Section 5.6.5 

do not affect the performance of any LSCS 

structure, system, or component credited 

with mitigating any accident previously 

evaluated. The NRC-approved analytical 

methodology for evaluating LOFWH 

transients will not affect the control 

parameters governing unit operation or the 

response of plant equipment to transient 

conditions. The proposed changes do not 

introduce any new modes of system 

operation or failure mechanism. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in the margin of safety? 

Response: No. 

The proposed changes will add a reference 

to the list of analytical methods in TS 5.6.5 

that can be used to determine core operating 

limits and delete seven obsolete references. 

The proposed changes do not modify the 

safety limits or setpoints at which protective 

actions are initiated and do not change the 

requirements governing operation or 

availability of safety equipment assumed to 

operate to preserve the margin of safety. 

Therefore, the proposed changes provide an 

equivalent level of protection as that 

currently provided. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

Based on the above information, EGC 

concludes that the proposed amendment 

presents no significant hazards consideration 

under the standards set forth in 10 CFR 

50.92(c), and, accordingly, a finding 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 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Thomas S. 

O’Neill, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Daniel S. Collins. 

Exelon Generation Company, LLC, and 

PSEG Nuclear LLC, Docket Nos. 50–277 

and 50–278, Peach Bottom Atomic 

Power Station, Units 2 and 3, York and 

Lancaster Counties, Pennsylvania 

Date of application for amendments: 

June 8, 2006. 

Description of amendment request: 

The proposed changes modify Technical 

Specifications (TSs) 3.1.3, ‘‘Control Rod 

OPERABILITY’’; 3.1.6, ‘‘Rod Pattern 

Control’’; 3.3.2.1, ‘‘Control Rod Block 

Instrumentation’’; 3.10.7, ‘‘Control Rod 

Testing—Operating’’; and 3.10.8, 

‘‘SHUTDOWN MARGIN (SDM) Test—

Refueling’’ to replace the current 

references to banked position 

withdrawal sequence (BPWS) with 

references to ‘‘the analyzed rod position 

sequence.’’

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change modifies Technical 

Specifications (TS) 3.1.3, ‘‘Control Rod 

OPERABILITY’’; TS 3.1.6, ‘‘Rod Pattern 

Control’’; TS 3.3.2.1, ‘‘Control Rod Block 

Instrumentation’’; TS 3.10.7, ‘‘Control Rod 

Testing—Operating’’, and; TS 3.10.8, 

SHUTDOWN MARGIN (SDM) Test—

Refueling’’. The proposed change would 

replace the current references to ‘‘Banked 

Position Withdrawal Sequence (BPWS)’’ with 

references to ‘‘the analyzed rod position 

sequence’’. The use of the ‘‘the analyzed rod 

position sequence’’ will continue to 

minimize the consequences of an accident 

previously evaluated including the Control 

Rod Drop Accident (CRDA). Additionally, 

the use of the words ‘‘the analyzed rod 

position sequence’’ will provide an 

equivalent level of protection during plant 

startups and shutdowns and therefore will 

not increase the consequences of an accident 

previously evaluated. 

Control rod patterns during startup and 

shutdown conditions will continue to be 

controlled by the operator and the Rod Worth 

Minimizer (RWM) (LCO [limiting condition 

of operation] 3.3.2.1, ‘‘Control Rod Block 

Instrumentation’’), so that only specified 

control rod sequences and relative positions 

are allowed over the operating range of all 

control rods inserted to 10% of Rated 

Thermal Power. As a result of this change, 

these sequences will continue to limit the 

potential amount of reactivity addition that 

could occur in the event of a Control Rod 

Drop Accident (CRDA). 

Accidents are initiated by the malfunction 

of plant equipment, or the failure of plant 

structures, systems, or components. The 

proposed change will ensure that analyzed 

rod position sequences are developed to 

minimize incremental control rod reactivity 

worth in accordance with the ‘‘General 

Electric Standard Application for Reactor 

Fuel,’’ NEDE–24011–P–A–15 (GESTAR–II), 

and U.S. Supplement, NEDE–24011–P–A–

15–US, September, 2005, NRC approved 

methodology, and reviewed and approved in 

accordance with the 10 CFR 50.59 process. 

These analyzed rod position sequences will 

limit the potential reactivity increase for a 

postulated CRDA during reactor startups and 

shutdowns below the Low Power Setpoint of 

10% of Rated Thermal Power. 

The proposed change will continue to 

ensure that systems, structures and 

components are capable of performing their 

intended safety functions. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not affect the 

assumed accident performance of the control 

rods, nor any plant structure, system, or 

component previously evaluated. 

The proposed change does not involve the 

installation of new equipment, and installed 

equipment is not being operated in a new or 

different manner. The change ensures that 

control rods remain capable of performing 

their safety functions. No set points are being 

changed which would alter the dynamic 

response of plant equipment. Accordingly, 

no new failure modes are introduced. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change will ensure that 

analyzed rod position sequences are 

developed to minimize incremental control 

rod reactivity worth in accordance with the 

‘‘General Electric Standard Application for 

Reactor Fuel,’’ NEDE–24011–P–A–15 

(GESTAR–II), and U.S. Supplement, NEDE–

24011–P–A–15–US, September, 2005, NRC 

approved methodology, and reviewed and 

approved in accordance with the 10 CFR 

50.59 process. The proposed change will not 

adversely impact the plant’s response to an 

accident or transient. All current safety 

margins will be maintained. There are no 

changes proposed which alter the set points 

at which protective actions are initiated, and 

there is no change to the operability 

requirements for equipment assumed to 

operate for accident mitigation. 

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 

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Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 46935 

amendment request involves no 

significant hazards consideration. 

Attorney for Licensee: Mr. Brad 

Fewell, Assistant General Counsel, 

Exelon Generation Company, LLC, 200 

Exelon Way, Kennett Square, PA 19348. 

NRC Branch Chief (Acting): Brooke D. 

Poole. 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

Station, Unit Nos. 1 and 2 (BVPS–1 

and 2), Beaver County, Pennsylvania 

Date of amendment request: June 14, 

2006. 

Description of amendment request: 

The amendments would incorporate the 

results of a new spent fuel pool 

criticality analysis documented in 

WCAP–16518–P/WCAP–16518–NP, 

‘‘Beaver Valley Unit 2 Spent Fuel Pool 

Criticality Analysis,’’ Revision 1, May 

2006 for the BVPS–2 spent fuel storage 

pool. The revised criticality analysis 

will permit utilization of vacant storage 

locations dictated by the existing 

Technical Specification (TS) storage 

configurations in the BVPS–2 spent fuel 

storage pool. 

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 relevant accidents 

previously evaluated are limited to the fuel 

handling and criticality accidents. 

Administrative controls during fuel 

fabrication ensure that the fuel is fabricated 

to ensure proper loading of fuel in the fuel 

assemblies. Administrative and operational 

controls used to load fuel assemblies into the 

spent fuel pool ensure the fuel assemblies are 

stored in compliance with the allowed 

storage configurations. Fuel handling is 

performed under administrative controls and 

physical limitations. These controls will 

remain in effect and continue to protect 

against criticality and fuel handling accidents 

involving new storage configurations dictated 

by the new analysis. There is therefore no 

impact on the probability of fuel handling or 

criticality accidents. 

The new criticality analysis defines new 

spent fuel storage configurations with new 

enrichment and burnup limits. Integral Fuel 

Burnable Absorber (IFBA) limits are used to 

comply with the 1-out-of-4 configuration for 

fresh fuel. The boron dilution evaluation that 

supported Amendment [No.] 128 [February 

11, 2002, Agencywide Documents Access 

and Management System Accession No. 

ML020020373], permitting credit for soluble 

boron at BVPS Unit No. 2 continues to 

remain valid. The new analysis demonstrates 

that keff remains below unity for the various 

storage configurations considered with zero 

soluble boron, and that keff remains less than 

or equal to 0.95 for the entire pool with credit 

for soluble boron under non-accident and 

accident conditions with a 95% probability 

at a 95% confidence level (95/95). Potential 

consequences of accidents previously 

analyzed remain unchanged. 

The editorial changes made to the table 

numbers and the LCO [Limiting Condition 

for Operation] and Surveillance Requirement 

wording do not impact probability or 

consequences of an accident previously 

evaluated. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. The relevant types of 

accidents previously evaluated are limited to 

criticality and fuel handling accidents. 

Although the new analysis will allow 

utilization of additional storage capacity, 

implementation of fuel loading 

configurations and fuel handling activities 

will continue to be performed under 

administrative and operational controls. No 

new or different activities are introduced as 

a result of the proposed changes. The 

utilization of additional storage capacity 

within the allowances of the revised analysis 

will introduce no new or other kind of 

accident. 

The editorial changes made to the table 

numbers and the LCO and Surveillance 

Requirement wording do not impact any 

previously evaluated accident. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. The margin to safety with 

respect to analyzed accidents involves 

maintaining keff through fuel storage 

configurations and boron concentration 

controls in the spent fuel pool. The boron 

dilution evaluation that supported that 

supported Amendment [No.] 128 permitting 

credit for soluble boron at BVPS Unit No. 2 

remains valid. The Amendment [No.] 128 

evaluation concluded that a boron dilution 

event is not credible for BVPS Unit No. 2. 

The new analysis calculates the non-accident 

soluble boron concentration to be less than 

was determined in the Amendment [No.] 128 

evaluation. Thus, there is no significant 

reduction in a margin of safety because of the 

new analysis and the conclusions of the 

Amendment [No.] 128 dilution evaluation 

remain valid. 

Under accident conditions, the soluble 

boron needed to maintain keff below 0.95 

with the new storage configurations is less 

than what is assumed in current analysis. 

The proposed change does not involve a 

significant reduction in a margin of safety for 

accident conditions. 

The editorial changes made to the table 

numbers and the LCO and Surveillance 

Requirement wording do not impact a margin 

of safety. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mary O’Reilly, 

FirstEnergy Nuclear Operating 

Company, FirstEnergy Corporation, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Richard J. Laufer. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit 1, Lake 

County, Ohio 

Date of amendment request: June 1, 

2006. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification 3.4.10, 

‘‘Residual Heat Removal (RHR) 

Shutdown Cooling System—Cold 

Shutdown’’ by adding a default 

Condition to address situations when an 

RHR shutdown cooling subsystem 

becomes inoperable in MODE 4 and, 

within the completion time of 1 hour, 

an alternate method of decay heat 

removal can not be verified to be 

available. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. The proposed amendment 

does not change the design of any structures, 

systems or components (SSCs), and does not 

affect the manner in which plant systems are 

operated. It is a change to the Technical 

Specifications only, to provide guidance to 

plant operators on appropriate actions to 

take, where no Technical Specification 

guidance currently exists. Since the design of 

plant SSCs is not changed and plant systems 

and components are not operated in a 

different manner, there is no change to 

previously identified accident initiators, and 

the proposed amendment would not impact 

the probability of any of the previously 

evaluated accidents in the Updated Safety 

Analysis Report (USAR). 

The USAR event that evaluates the 

consequences of a loss of RHR Shutdown 

Cooling is included in Section 15.2.9 entitled 

‘‘Failure of RHR Shutdown Cooling’’. This 

event examines the consequences of a loss of 

not only an RHR shutdown cooling 

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46936 Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 

subsystem, but also the loss of the suction 

source from the recirculation system leading 

to both RHR Shutdown Cooling subsystems, 

and a loss of offsite power. Even with these 

multiple failures, this event is not one of the 

limiting transients. As noted in Section 

15.2.9.5, ‘‘Radiological Consequences,’’ there 

are no fuel failures, and the consequences of 

the event are much less than those for the 

‘‘Main Steam Isolation Valve Closure’’

transient, which is evaluated with acceptable 

results in USAR Section 15.2.4.5. Since the 

proposed amendment only involves the 

addition of a Required Action where no 

guidance currently exists, and the design of 

plant SSCs is not changed and plant systems 

and components are not operated in a 

different manner, the proposed amendment 

does not affect the consequences of the 

Section 15.2.9 analysis, nor does it affect the 

ability of the installed RHR subsystems to 

perform their shutdown cooling function. 

The change adds a default Condition to 

provide guidance to the operators in those 

situations when a subsystem becomes 

inoperable with the plant in MODE 4 and an 

alternate cannot be verified to be available 

within an hour, which does not impact the 

consequences of the previously evaluated 

accidents in the USAR. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. This change to the required 

Technical Specification actions does not 

involve a change in the design function or 

operation of plant SSCs. It does not introduce 

credible new failure mechanisms, 

malfunctions, or accident initiators not 

considered in the existing plant design and 

licensing basis. 

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 amendment involve 

a significant reduction in a margin of safety? 

Response: No. This proposed amendment 

only involves a change to the required 

Technical Specification actions. It does not 

involve a change in the evaluation and 

analysis methods used to demonstrate 

compliance with regulatory and licensing 

requirements, and does not exceed or alter a 

design basis or safety limit. The safety margin 

before the change remains unchanged after 

the proposed amendment. 

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: David W. 

Jenkins, Attorney, FirstEnergy 

Corporation, 76 South Main Street, 

Akron, OH 44308. 

NRC Branch Chief: Daniel S. Collins. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit 1, Lake 

County, Ohio 

Date of amendment request: June 1, 

2006. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification 3.4.9, 

‘‘Residual Heat Removal (RHR) 

Shutdown Cooling System—Hot 

Shutdown,’’ to revise the Required 

Actions when both RHR shutdown 

cooling subsystems are inoperable in 

MODE 3. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. The proposed amendment 

does not change the design of any structures, 

systems or components (SSCs), and does not 

affect the manner in which plant systems are 

operated. It is a change to the Technical 

Specifications only, to provide guidance to 

plant operators on appropriate actions to 

take, when both RHR shutdown cooling 

subsystems are inoperable. Since the design 

of plant SSCs is not changed and plant 

systems and components are not operated in 

a different manner, there is no change to 

previously identified accident initiators, and 

the proposed amendment would not impact 

the probability of any of the previously 

evaluated accidents in the Updated Safety 

Analysis Report (USAR). 

The USAR event that evaluates the 

consequences of a loss of RHR Shutdown 

Cooling is included in Section 15.2.9 entitled 

‘‘Failure of RHR Shutdown Cooling.’’ This 

event examines the consequences of a loss of 

not only an RHR shutdown cooling 

subsystem, but also the loss of the suction 

source from the recirculation system leading 

to both RHR Shutdown Cooling subsystems, 

and a loss of offsite power. Even with these 

multiple failures, this event is not one of the 

limiting transients. As noted in Section 

15.2.9.5, ‘‘Radiological Consequences,’’ there 

are no fuel failures, and the consequences of 

the event are much less than those for the 

‘‘Main Steam Isolation Valve Closure’’

transient, which is evaluated with acceptable 

results in USAR Section 15.2.4.5. Since the 

proposed amendment only involves the 

addition of a Required Action where no 

guidance currently exists, and the design of 

plant SSCs is not changed and plant systems 

and components are not operated in a 

different manner, the proposed amendment 

does not affect the consequences of the 

Section 15.2.9 analysis, nor does it affect the 

ability of the installed RHR subsystems to 

perform their shutdown cooling function. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. This change to the required 

Technical Specification actions does not 

involve a change in the design function or 

operation of plant SSCs. It does not introduce 

credible new failure mechanisms, 

malfunctions, or accident initiators not 

considered in the existing plant design and 

licensing basis. 

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 amendment involve 

a significant reduction in a margin of safety? 

Response: No. This proposed amendment 

only involves a change to the required 

Technical Specification actions. It does not 

involve a change in the evaluation and 

analysis methods used to demonstrate 

compliance with regulatory and licensing 

requirements, and does not exceed or alter a 

design basis or safety limit. The safety margin 

before the change remains unchanged after 

the proposed amendment. 

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: David W. 

Jenkins, Attorney, FirstEnergy 

Corporation, 76 South Main Street, 

Akron, OH 44308. 

NRC Branch Chief: Daniel S. Collins. 

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 amendment request: April 28, 

2006. 

Description of amendment request: 

The proposed amendment would 

change the SSES 1 and 2 Technical 

Specifications (TSs) to modify the 

standby liquid control system for single 

loop pump operation and use of 

enriched sodium pentaborate solution. 

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: 

Does the proposed change involve a 

significant increase in the probability or 

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consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes revise Technical 

Specification 3.1.7 for the Standby Liquid 

Control (SLC) system to reflect new boron 

weight-percent and enrichment 

requirements. In addition, the change to 

single pump operation reduces the required 

SLC pump flow and discharge pressure 

required to satisfy 10 CFR 50.62, thus 

increasing the reliability of the system. The 

changes do not otherwise alter the design or 

operation of the SLC system, and the existing 

design of the system is sufficient to support 

operation with the enriched sodium 

pentaborate solution. The SLC system is not 

considered to be the initiator of any event 

currently analyzed in the FSAR [Final Safety 

Analysis Report]. Therefore, the proposed 

changes do not increase the probability of a 

previously evaluated accident. 

The SSES ATWS [anticipated transient 

without scram] analysis was performed using 

standard accepted assumptions, inputs, and 

codes. That analysis, which demonstrated 

that the acceptance criteria for peak vessel 

pressure, peak cladding temperature, peak 

local cladding oxidation, peak suppression 

pool temperature, and peak containment 

pressure, established the requirements for the 

proposed boron weight-percent and 

concentration, and pump flow rate. The 

analysis assumed the use of only a single 

pump, versus two pumps. The results of the 

analysis are that no fission product barriers 

are adversely challenged, and the 

radiological consequences of previously 

evaluated accidents (i.e., ATWS) are not 

increased. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes revise Technical 

Specification 3.1.7 for the SLC system to 

reflect new boron weight-percent and 

enrichment requirements. In addition, the 

change to single pump operation reduces the 

required SLC pump flow and discharge 

pressure required to satisfy 10 CFR 50.62, 

thus increasing the reliability of the system. 

A new Surveillance Requirement (SR 

3.1.7.10) is also added to verify the correct 

solution enrichment prior to addition of 

inventory to the SLC tank. The changes do 

not otherwise alter the design or operation of 

the SLC system, and the existing design of 

the system is sufficient to process the 

enriched sodium pentaborate solution. With 

the exception of these changes, no other 

physical changes to plant structures or 

systems are proposed. Thus, the proposed 

changes do not create a new initiating event 

for the spectrum of events currently 

postulated in the FSAR. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes revise Technical 

Specification 3.1.7 for the SLC system to 

reflect new boron weight-percent and 

enrichment requirements. In addition, the 

change to single pump operation reduces the 

required SLC pump flow and discharge 

pressure required to satisfy 10 CFR 50.62, 

thus increasing the reliability of the system. 

The changes do not otherwise alter the 

design or operation of the SLC system, and 

the existing design of the system is sufficient 

to process the enriched sodium pentaborate 

solution. 

The analysis was performed using standard 

accepted assumptions, inputs, and codes. 

That analysis, which demonstrated that 

ATWS acceptance criteria are satisfied, 

established the requirements for the 

proposed boron weight-percent and 

concentration, and pump flow rate. Further, 

the analysis assumed only a single pump is 

in operation verses two pumps. The 

evaluation demonstrated that the SLC system 

meets this post-LOCA [loss-of-coolant 

accident] suppression pool pH control design 

function. 

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: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179. 

NRC Branch Chief: Richard J. Laufer. 

Tennessee Valley Authority, Docket No. 

50–259 , Browns Ferry Nuclear Plant, 

Unit 1, Limestone County, Alabama 

Date of amendment request: October 

12, 2004. 

Description of amendment request: As 

part of Nuclear Regulatory 

Commission’s (NRC) approval of the 

Improved Technical Specifications for 

Browns Ferry Nuclear Plant, Unit 1, by 

Amendment No. 234, NRC imposed 

License Condition 2.C(4) to ensure that 

the required analyses and modifications 

needed to support the Technical 

Specification (TS) changes made by 

License Amendment No. 234 and any 

subsequent TS changes, were completed 

by licensee prior to entering the mode 

for which the TS applies. The proposed 

amendment would remove this license 

condition from the license. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment does not affect 

any precursors for accidents described in 

Chapter 14 of the Browns Ferry Updated 

Final Safety Analysis Report (UFSAR). The 

proposed amendment does not change the 

conditions, operating configurations, or 

minimum amount of operating equipment 

assumed in the safety analysis for accident 

mitigation. No changes are proposed in plant 

protection or which create new modes of 

plant operation. Therefore, the proposed 

amendment does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment does not 

introduce new equipment, which could 

create a new or different kind of accident. No 

new external threats, release pathways, or 

equipment failure modes are created. 

Therefore, the implementation of the 

proposed amendment will not create a 

possibility for an accident of a new or 

different type than those previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed amendment does not impact 

the redundancy or availability of equipment 

credited in the response to accidents 

described in Chapter 14 of the UFSAR. For 

these reasons, the proposed amendment does 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: L. Raghavan. 

Tennessee Valley Authority, Docket No. 

50–259, Browns Ferry Nuclear Plant, 

Unit 1, Limestone County, Alabama 

Date of amendment request: May 1, 

2006 (TS–455). 

Description of amendment request: 

The proposed amendment would revise 

the numeric values of the safety limit 

minimum critical power ratio (SLMCPR) 

in the Technical Specification (TS) 

Section 2.1.1.2 for single and two 

reactor recirculation loop operation to 

incorporate the results of the Browns 

Ferry Nuclear Plant, Unit 1 Cycle 7 

SLMCPR analysis. 

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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 Technical 

Specification change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

Response: No. 

The proposed amendment establishes a 

revised SLMCPR value for single and two 

recirculation loop operation. The probability 

of an evaluated accident is derived from the 

probabilities of the individual precursors to 

that accident. The proposed SLMCPR values 

preserve the existing margin to transition 

boiling and the probability of fuel damage is 

not increased. Since the change does not 

require any physical plant modifications or 

physically affect any plant components, no 

individual precursors of an accident are 

affected and the probability of an evaluated 

accident is not increased by revising the 

SLMCPR values. 

The consequences of an evaluated accident 

are determined by the operability of plant 

systems designed to mitigate those 

consequences. The revised SLMCPR values 

have been determined using NRC-approved 

methods and procedures. The basis of the 

MCPR Safety Limit is to ensure no 

mechanistic fuel damage is calculated to 

occur if the limit is not violated. These 

calculations do not change the method of 

operating the plant and have no effect on the 

consequences of an evaluated accident. 

Therefore, the proposed TS change does not 

involve an increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed Technical 

Specification change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed license amendment involves 

a revision of the SLMCPR value for single 

and two recirculation loop operation based 

on the results of an analysis of the Unit 1 

Cycle 7 core. Creation of the possibility of a 

new or different kind of accident would 

require the creation of one or more new 

precursors of that accident. New accident 

precursors may be created by modifications 

of the plant configuration, including changes 

in the allowable methods of operating the 

facility. This proposed license amendment 

does not involve any modifications of the 

plant configuration or changes in the 

allowable methods of operation. Therefore, 

the proposed TS change does not create the 

possibility of a new or different kind of 

accident previously evaluated. 

3. Does the proposed Technical 

Specification change involve a significant 

reduction in a margin of safety? 

Response: No. 

The margin of safety as defined in the TS 

bases will remain the same. The new 

SLMCPR values were calculated using 

referenced fuel vendor methods and 

procedures, which are in accordance with the 

fuel design and licensing criteria. The 

SLMCPR remains high enough to ensure that 

greater than 99.9 percent of all fuel rods in 

the core are expected to avoid transition 

boiling if the limit is not violated, thereby 

preserving the fuel cladding integrity. 

Therefore, the proposed TS change does not 

involve a reduction in the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC (Acting) Branch Chief: L. 

Raghavan. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant, Units 1 and 2, Hamilton 

County, Tennessee 

Date of amendment request: July 6, 

2006 (TS–06–04). 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specifications (TS) for the 

Sequoyah Nuclear Plant, Units 1 and 2. 

Action a.1 of TS 3.1.3.2, ‘‘Position 

Indication Systems—Operating,’’

requires the verification of rod position 

by use of the moveable incore detectors. 

Tennessee Valley Authority (the 

licensee, TVA) is proposing a revision to 

TS 3.1.3.2 to allow the position of the 

control and shutdown rods to be 

monitored by a means other than the 

moveable incore detectors. The 

amendment will provide a less 

burdensome monitoring method should 

problems with the analog rod position 

indication (ARPI) system be 

experienced. When a recurring problem 

in the system requires the monitoring of 

a rod’s position by the alternate means, 

TVA plans to continue unit operation 

and to use the alternate means until the 

unit enters Mode 5 and repairs to the 

system can safely be implemented. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change provides an 

alternative method for the monitoring of the 

position of a rod once the position of the rod 

is verified using the moveable incore detector 

system. The proposed monitoring of rod 

control system parameters provides a 

reasonably similar approach to rod position 

monitoring as that provided by the movable 

incore detector system. In particular, the 

ability to immediately detect a rod drop or 

misalignment is not directly provided by the 

movable incore detector system or by the 

monitoring of rod control system parameters. 

Additionally, neither the movable incore 

detector system, nor the monitoring of rod 

control system parameters, provides the 

capability to verify rod position following a 

reactor trip or shutdown. Therefore, the 

monitoring of rod control system parameters, 

in lieu of the use of the movable incore 

detector system, provides an equivalent and 

acceptable method of monitoring rod 

position while a position indicator is 

inoperable. 

The proposed change does not alter plant 

equipment that is considered to have the 

potential to alter the probability of an 

accident. The affected components are for 

monitoring only and do not actively affect 

equipment that interacts with the control of 

the reactor. Likewise, the affected 

components are for monitoring and provide 

an equivalent level of indication of rod 

position as the current action. This maintains 

an acceptable level of rod position indication 

for normal plant operations, as well as post 

accident mitigation actions. 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. 

As described above, the proposed change 

provides only an alternative method of 

monitoring the position of a rod. No new 

accident initiators are introduced by the 

proposed alternative manner of performing 

rod position monitoring. The proposed 

change does not affect the reactor protection 

system or the reactor control system. Hence, 

no new failure modes are created that would 

cause a new or different kind of accident 

from any accident previously evaluated. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The rod position indicators are required to 

determine control rod positions and thereby 

ensure compliance with the control rod 

alignment and insertion limits. The proposed 

change does not alter the requirement to 

determine rod position but provides an 

alternative method for monitoring the 

position of the affected rod after the position 

of the rod is verified using the moveable 

incore detector system. As a result, the initial 

conditions of the accident analysis are 

preserved. The components affected by the 

alternate rod monitoring will not affect plant 

setpoints utilized for automatic mitigation of 

accident conditions or other equipment 

necessary for accident mitigation. 

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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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: Michael L. 

Marshall, Jr. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant, Units 1 and 2, Hamilton 

County, Tennessee 

Date of amendment request: July 12, 

2006 (TS–06–03). 

Description of amendment request: 

The proposed amendment would revise 

the limiting condition for operation for 

the Sequoyah Nuclear Plant, Units 1 and 

2, Technical Specification (TS) Section 

3.7.5, ‘‘Ultimate Heat Sink.’’ This 

revision would change the minimum 

ultimate heat sink (UHS) water 

elevation in TS 3.7.5.a from 670 feet to 

674 feet. The essential raw cooling 

water (ERCW) temperature requirement 

in TS 3.7.5.b would be increased from 

83 degrees Fahrenheit (°F) to 87 °F. The 

conditional requirements of TS 3.7.5.c 

would no longer be required and would 

be deleted by the proposed change. This 

change would also delete a footnote that 

established a temporary UHS 

temperature limit of 87 °F through 

September 30, 1995. These proposed 

changes are supported by a combination 

of design basis re-analysis, bounding 

analysis, and sensitivity analysis of the 

ERCW system, the UHS, and supported 

systems. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change to increase the UHS 

maximum temperature and the minimum 

water level does not alter the function, 

design, or operating practices for plant 

systems or components. One exception is the 

elimination of non-safety-related station air 

compressor loads located in the turbine 

building. The UHS is utilized to remove heat 

loads from plant systems during normal and 

accident conditions. This function is not 

expected or postulated to result in the 

generation of any accident and continues to 

adequately satisfy the associated safety 

functions with the proposed changes. 

Therefore, the probability of an accident 

presently evaluated in the safety analyses 

will not be increased because the UHS 

function does not have the potential to be the 

source of an accident. The heat loads that the 

UHS is designed to accommodate have been 

evaluated for functionality with the higher 

temperature and elevation requirements. The 

result of these evaluations is that there is 

existing margins associated with the systems 

that utilize the UHS for normal and accident 

conditions. These margins are sufficient to 

accommodate the postulated normal and 

accident heat loads with the proposed 

changes to the UHS. Since the safety 

functions of the UHS are maintained, the 

systems that ensure acceptable offsite dose 

consequences will continue to operate as 

designed. 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 UHS function is not an initiator of any 

accident and only serves as a heat sink for 

normal and upset plant conditions. By 

allowing the proposed change in the UHS 

temperature and elevation requirements, only 

the parameters for UHS operation are 

changed while the safety functions of the 

UHS and systems that transfer the heat sink 

capability continue to be maintained. The 

UHS function provides accident mitigation 

capabilities and does not reflect the potential 

for accident generation. Therefore, the 

possibility for creating a new or different 

kind of accident is not created because the 

UHS is only utilized for heat removal 

functions that are not a potential source for 

accident generation. Therefore, the proposed 

change does not create the possibility of a 

new or different kind of accident from any 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change has been evaluated 

for systems that are needed to support 

accident mitigation functions as well as 

normal operational evolutions. Operational 

margins were found to exist in the systems 

that utilize the UHS capabilities such that 

these proposed changes will not result in the 

loss of any safety function necessary for 

normal or accident conditions. The ERCW 

system has excess flow margins that will 

accommodate the increased flows necessary 

for the proposed temperature increase. While 

operating margins have been reduced by the 

proposed changes, safety margins have been 

maintained as assumed in the accident 

analyses for postulated events. 

Additionally, the proposed changes do not 

require the modification of component 

setpoints utilized for automatic mitigation of 

accident conditions or other equipment 

necessary for accident mitigation. Therefore, 

a significant reduction in the margin to safety 

is not created by this proposed change. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: L. Raghavan. 

Tennessee Valley Authority, Docket No. 

50–390, Watts Bar Nuclear Plant, Unit 1, 

Rhea County, Tennessee 

Date of amendment request: June 16, 

2006 (WBN–TS–06–04). 

Description of amendment request: 

The proposed amendment change 

would revise Technical Specification 

(TS) 5.7.2.11, ‘‘Inservice Testing 

Program,’’ to remove ‘‘applicable 

supports’’ from the Inservice Testing 

(IST) Program and revise the IST 

Program for pumps and valves to meet 

the requirements of the latest Edition 

and Addenda of the American Society 

of Mechanical Engineers (ASME) Code 

approved by the NRC for use on the date 

12-months prior to the start of the 10- 

year IST Interval. For the Watts Bar 

Nuclear Plant (WBN), Unit 1, the second 

10-year IST Interval will begin on 

December 27, 2006. The ASME Code 

that was approved in 10 CFR 

50.55a(f)(4) for use on December 27, 

2005, was ASME Operations and 

Maintenance (OM) Code, 2001 Edition, 

with Addenda through 2003. The 

proposed change provides consistency 

with the requirements in 10 CFR 

50.55a(f)(4) by replacing the reference to 

ASME Boiler and Pressure Vessel Code, 

Section XI, with ASME OM Code. This 

proposed change is based on Technical 

Specification Task Force (TSTF) 

Traveler 479, Revision 0, ‘‘Changes to 

Reflect Revision of 10 CFR 50.55a.’’

TSTF 279–A, Revision 0, ‘‘Remove 

‘applicable supports’ from Inservice 

Testing Program,’’ was approved by 

NRC and incorporated into Revision 2 of 

NUREG–1431, ‘‘Standard Technical 

Specification Westinghouse Plants.’’ In 

addition, the proposed amendment 

would add provisions to TS 5.7.2.11, 

Item b, to only apply Surveillance 

Requirement 3.0.2 to those IST 

frequencies of 2 years or less. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

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licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Technical 

Specification Section 5.7.2.11 for WBN Unit 

1 to conform to the requirements of 10 CFR 

50.55a(f)(4) regarding the inservice testing of 

pumps and valves which are classified as 

ASME Code Class 1, 2, and 3. 

ASME has in the last several years, 

transitioned the requirements for inservice 

testing of pumps and valves out of ASME 

Section XI and into a separate, stand alone 

code entitled the ‘‘Code for Operation and 

Maintenance of Nuclear Power Plants,’’

(ASME OM Code). The ASME OM Code has 

been endorsed by the NRC in 10 CFR 50.55a 

and is the Code that will be required for 

inservice testing of pumps and valves during 

the WBN Second Inservice Interval. The 

proposed change incorporates revisions to 

the ASME Code that result in a net 

improvement in the measures for testing 

pumps and valves. The proposed change also 

deletes the reference to supports from the 

Inservice Testing Program as supports are 

already inspected under the Inservice 

Inspection Program. 

The proposed changes do not involve any 

hardware changes, nor do the changes affect 

the probability of any event initiators. There 

will be no change to normal plant operating 

parameters, accident mitigation capabilities, 

or accident analysis assumptions or inputs. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change revises the Technical 

Specifications to delete the reference to 

‘‘applicable supports’’ from the Inservice 

Testing Program and to incorporate the latest 

Code requirements in 10 CFR 50.55a(f)(4) for 

Code Class 1, 2, and 3 pumps and valves for 

WBN’s next ten year interval. The testing 

requirements are similar and reflect the same 

type testing. Valves are still stroke timed; 

remote position indicators are still verified to 

be accurate; seat leakage measurements of 

critical valves are still performed; relief 

valves still have their setpoints and seat 

leakages verified; pumps are still tested for 

hydraulic performance and mechanical 

condition; check valves are verified to open 

and close properly; and supports are still 

inspected under the appropriate inspection 

program. 

The proposed changes do not involve a 

modification to the physical configuration of 

the plant or change methods governing 

normal plant operation. No test methods are 

added or deleted. Therefore, the proposed 

change does not create the possibility of a 

new or different kind of accident from any 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change revises the TS for 

consistency with the Standard Technical 

Specification and with the requirements in 

10 CFR 50.55a(f)(4) regarding the inservice 

testing of pumps and valves which are 

classified as ASME Code Class 1, 2, and 3. 

This change incorporates revisions to the 

ASME Code that result in a net improvement 

in the measures of testing. Incorporation of 

the ASME OM Code does not alter the 

limiting values and acceptance criteria used 

to judge the continued acceptability of 

components tested by the Inservice Testing 

Program. Deletion of the reference to 

supports in the Inservice Testing Program 

does not alter the support inspection program 

as the program is currently under the 

Inservice Inspection Program. Since these 

limits are not altered, the margin of safety is 

not altered. 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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: L. Raghavan. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: May 30, 

2006. 

Description of amendment request: 

The amendment would revise 

Surveillance Requirements (SRs) 3.5.2.8 

and 3.6.7.1 in the Technical 

Specifications (TSs), and delete the 

footnote to the frequency for SR 3.5.2.5. 

SR 3.5.2.8 would be revised by 

replacing the phrase ‘‘trash racks and 

screens’’ with the word ‘‘strainers.’’ This 

reflects (1) the replacement of the 

existing containment recirculation sump 

suction inlet trash racks and screens 

with strainers with significantly greater 

effective surface area, and (2) the 

resulting relocation of the recirculation 

fluid pH control system in Refueling 

Outage 15 schedule for the spring of 

2007. The footnote to SR 3.5.2.5 would 

be deleted because it is no longer 

applicable to the TSs. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do[es] the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

None of the changes impact the initiation 

or probability of occurrence of any accident 

[previously evaluated]. 

The consequences of accidents evaluated 

in the FSAR [Final Safety Analysis Report for 

the Callaway Plant] that could be affected by 

this proposed change are those involving the 

pressurization of the containment and 

associated flooding of the containment and 

recirculation of this fluid within the 

Emergency Core Cooling System (ECCS) or 

the Containment Spray System (CSS) (e.g., 

LOCAs [Loss-of-Coolant Accidents]). [The 

containment sump trash racks and screens, 

and the sump strainers that are replacing the 

trash racks and screens are not initiators of 

accidents.] 

Although the configurations of the existing 

sump screen and the replacement strainer 

assemblies are different, they serve the same 

fundamental purpose of passively removing 

debris from the suction of the supported 

system pumps. Removal of trash racks does 

not impact the adequacy of the pump NPSH 

[net positive suction head] assumed in the 

safety analyses. Likewise the change does not 

reduce the reliability of any supported 

systems or introduce any new system 

interactions. The greatly increased surface 

area of the new strainer is designed to reduce 

head loss [at the containment sump] and 

reduce the approach velocity at the strainer 

face significantly, decreasing the risk of 

impact from large debris entrained in the 

sump flow stream. 

The recirculation fluid pH control system 

storage baskets serve a passive function to 

provide a buffering agent to neutralize the 

sump solution. The redesign and relocation 

of the storage baskets are considered a like 

kind replacement. The baskets will be 

located within the flood plain and will 

continue to ensure that the buffering agent is 

dissolved in the sump fluid to ensure an 

equilibrium pH ≥ 7.1. Failure of a basket 

would not initiate an accident. The ECCS and 

CSS will continue to function in a manner 

consistent with the plant design basis. 

As such, the proposed change to the 

Technical Specifications Surveillance 

Requirements does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. The 

installed quantity of trisodium phosphate 

Crystalline will provide a minimum 

equilibrium sump pH of 7.1 following 

dissolution and mixing. [Deleting the 

footnote to SR 3.5.2.5 is an administrative 

change to remove a one-time required 

verification that has already been performed 

and is no longer a requirement in the current 

TSs.] Therefore, there is not a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

2. Do[es] the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The containment recirculation sump 

strainers and recirculation fluid pH control 

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Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 46941 

system are passive systems used for accident 

mitigation. As such, they cannot be accident 

initiators. Therefore, there is no possibility 

that this change could create any accident of 

any kind. [The containment recirculation 

sump suction inlet trash racks and screens 

are being replaced with a complex strainer 

design with significantly larger effective 

surface area to reduce head loss and reduce 

the approach velocity at the strainer face 

significantly, decreasing the risk of impact 

from large debris entrained in the sump flow 

stream. This will result in the recirculation 

fluid pH control system being relocated.] 

No new accident scenarios, transient 

precursors, or limiting single failures are 

introduced as a result of these changes. There 

will be no adverse effect[s] or challenges 

imposed on any safety-related system as a 

result of these changes. The quantity of 

trisodium phosphate crystalline will provide 

a minimum equilibrium sump pH of ≥ 7.1 

following dissolution and mixing. Therefore, 

the possibility of a new or different type of 

accident is not created. 

There are no changes which would cause 

the malfunction of safety-related equipment, 

assumed to be operable in the accident 

analyses, as a result of the proposed 

Technical Specification changes. No new 

equipment performance burdens are 

imposed. The possibility of a malfunction of 

safety-related equipment with a different 

result is not created. [Deleting the footnote to 

SR 3.5.2.5 is an administrative change to 

remove a one-time required verification that 

has already been performed and is no longer 

a requirement in the current TSs.] Therefore, 

the proposed change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Do[es] the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not adversely 

affect any plant safety limits, setpoints, or 

design parameters. The changes also do not 

adversely affect the fuel, fuel cladding, 

Reactor Coolant System (RCS), or 

containment integrity. [The radiological dose 

consequence acceptance criteria in the 

Standard Review Plan for accidents will 

continue to be met. Deleting the footnote to 

SR 3.5.2.5 is an administrative change to 

remove a one-time required verification that 

has already been performed and is no longer 

a requirement in the current TSs.] Therefore, 

the proposed TS 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: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street, NW., Washington, 

DC 20037. 

NRC Branch Chief: David Terao. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Units No. 1 and 

No. 2, Louisa County, Virginia 

Date of amendment request: May 30, 

2006, as supplemented by letter dated 

June 30, 2006. 

Description of amendment request: 

The proposed amendments would 

relocate the American Society for 

Testing and Materials (ASTM) standard 

being used to test the total particulate 

concentration of the stored fuel oil to 

the TS Bases. This proposed change is 

described in TS Task force (TSTF) 

Standard TS Change Traveler TSTF–

374–A, Rev. 0, ‘‘Revision to TS 5.5.13 

and Associated TS Bases for Diesel Fuel 

Oil.’’ In addition, the licensee has 

proposed to use a ‘‘water and sediment 

test’’ instead of the ‘‘clear and bright’’

test provided in TSTF–374. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do changes involve a significant increase 

in the probability or consequences of an 

accident previously evaluated? 

The proposed change relocates the specific 

ASTM reference from the Administrative 

Controls Section of Technical Specifications 

(TS) to a licensee-controlled document. 

Relocating the specific ASTM Standard 

reference from the TS to a licensee-controlled 

document will not affect nor degrade the 

ability of the EDGs [emergency diesel 

generators] to perform their specified safety 

function. Fuel oil quality will continue to 

meet the current ASTM requirements for 

particulate concentration. 

The proposed change is administrative in 

nature and does not adversely affect accident 

initiators or precursors nor alter the design 

assumptions, conditions, and configuration 

of the facility or the manner in which the 

plant is operated and maintained. The 

proposed change does not alter or prevent the 

ability of structures, systems or components 

from performing their intended function to 

mitigate the consequences on an initiating 

event with the assumed acceptance limits. 

The proposed change does not affect the 

source term, containment isolation, or 

radiological release assumptions used in 

evaluating the radiological consequences of 

an accident previously evaluated. Further, 

the proposed change does not increase the 

types and amounts of radioactive effluent 

that may be released offsite, nor significantly 

increase individual or cumulative 

occupational/public radiation exposure. 

Therefore, the change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Do changes create the possibility of a 

new or different kind of accident from any 

accident previously evaluated? 

The proposed change relocates the specific 

ASTM reference from the Administrative 

Controls Section of Technical Specifications 

to a licensee-controlled document. 

The change does not involve a physical 

alteration of the plant or a change in the 

methods governing normal plant conditions. 

In addition, the change does not impose any 

new or different requirements or eliminate 

any existing requirements. The change does 

not alter assumptions made in the safety 

analysis and licensing basis. Therefore, the 

change does not create the possibility of a 

new or different kind of accident from any 

accident previously evaluated. 

3. Do changes involve a significant 

reduction in the margin of safety? 

The proposed change relocates the specific 

ASTM reference from the Administrative 

Controls Section of TS to a licenseecontrolled document. The detail associated 

with the specific ASTM standard reference is 

not required to be in the TS to provide 

adequate protection of the public health and 

safety, since the TS still retain the 

requirement for compliance with the 

applicable ASTM standard. 

The level of safety of facility operation is 

unaffected by the proposed change since 

there is no change in the intent of the TS 

requirements of assuring fuel oil is of the 

appropriate quality for EDG use. The 

proposed change provides the flexibility 

needed to maintain state-of-the-art 

technology in fuel oil sampling and analysis 

methodology. 

The proposed change does not reduce a 

margin of safety since it has no impact on 

any transient or 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: Lillian M. 

Cuoco, Esq., Senior Counsel, Dominion 

Resources Services, Inc., Millstone 

Power Station, Building 475, 5th Floor, 

Rope Ferry Road, Rt. 156, Waterford, 

Connecticut 06385. 

NRC Branch Chief: Evangelos C. 

Marinos. 

Virginia Electric and Power Company, 

Docket Nos. 50–280 and 50–281, Surry 

Power Station, Unit Nos. 1 and 2, Surry 

County, Virginia 

Date of amendment request: May 26, 

2006. 

Description of amendment request: 

Item 1: The proposed amendments 

would revise the Technical 

Specification (TS) requirements related 

to Reactor Coolant System (RCS) leakage 

definitions and requirements and steam 

generator tube integrity. The licensee 

requested this change to implement TS 

Task Force (TSTF) Standard TS Change 

Traveler, TSTF–449, ‘‘Steam Generator 

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46942 Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 

Tube Integrity,’’ (TSTF–449, Rev. 4). 

Item 2: In addition, in its submittal 

dated May 26, 2006, the licensee 

proposed minor deviations from the TS 

changes described in TSTF–449, Rev. 4, 

to provide consistency with Surry’s 

custom TSs. 

Basis for proposed no significant 

hazards consideration determination: 

Item 1: As required by 10 CFR 50.91(a), 

an analysis of the issue of no significant 

hazards consideration is presented 

below: 

1. The Proposed Change Does Not Involve 

a Significant Increase in the Probability or 

Consequences of an Accident Previously 

Evaluated. 

The proposed change requires a SG 

Program that includes performance criteria 

that will provide reasonable assurance that 

the SG tubing will retain integrity over the 

full range of operating conditions (including 

startup, operation in the power range, hot 

standby, cooldown and all anticipated 

transients included in the design 

specification). The SG performance criteria 

are based on tube structural integrity, 

accident induced leakage, and operational 

leakage. 

A SG tube rupture (TR) event is one of the 

design basis accidents that are analyzed as 

part of a plant’s licensing basis. In the 

analysis of a SGTR event, a bounding 

primary to secondary leakage rate equal to 

the operational leakage rate limits in the 

licensing basis plus the leakage rate 

associated with a double-ended rupture of a 

single tube is assumed. 

For other design basis accidents such as 

main steam line break (MSLB), rod ejection, 

and reactor coolant pump locked rotor the 

tubes are assumed to retain their structural 

integrity (i.e., they are assumed not to 

rupture). These analyses typically assume 

that primary to secondary leakage for all SGs 

is 1 gallon per minute or increases to 1 gallon 

per minute as a result of accident induced 

stresses. The accident induced leakage 

criterion introduced by the proposed changes 

accounts for tubes that may leak during 

design basis accidents. The accident induced 

leakage criterion limits this leakage to no 

more than the value assumed in the accident 

analysis. 

The SG performance criteria proposed 

change to the TS identify the standards 

against which tube integrity is to be 

measured. Meeting the performance criteria 

provides reasonable assurance that the SG 

tubing will remain capable of fulfilling its 

specific safety function of maintaining 

reactor coolant pressure boundary integrity 

throughout each operating cycle and in the 

unlikely event of a design basis accident. The 

performance criteria are only a part of the SG 

Program required by the proposed change to 

the TS. The program, defined by NEI 97–06, 

Steam Generator Program Guidelines, 

includes a framework that incorporates a 

balance of prevention, inspection, evaluation, 

repair, and leakage monitoring. The proposed 

changes do not, therefore, significantly 

increase the probability of an accident 

previously evaluated. 

The consequences of design basis accidents 

are, in part, functions of the DOSE 

EQUIVALENT 1–131 in the primary coolant 

and the primary to secondary leakage rates 

resulting from an accident. Therefore, limits 

are included in the plant TS for operational 

leakage and for DOSE EQUIVALENT 1–131 

in primary coolant to ensure the plant is 

operated within its analyzed condition. The 

typical analysis of the limiting design basis 

accident assumes that primary to secondary 

leak rate after the accident is 1 gallon per 

minute with no more than 500 gallons per 

day in any one SG, and that the reactor 

coolant activity levels of DOSE 

EQUIVALENT 1–131 are at the TS values 

before the accident. 

The proposed change does not affect the 

design of the SGs, their method of operation, 

or primary coolant chemistry controls. The 

proposed approach updates the current TSs 

and enhances the requirements for SG 

inspections. The proposed change does not 

adversely impact any other previously 

evaluated design basis accident and is an 

improvement over the current TSs. 

Therefore, the proposed change does not 

affect the consequences of a SGTR accident 

and the probability of such an accident is 

reduced. In addition, the proposed changes 

do not affect the consequences of an MSLB, 

rod ejection, or a reactor coolant pump 

locked rotor event, or other previously 

evaluated accident. 

2. The Proposed Change Does Not Create 

the Possibility of a New or Different Kind of 

Accident from any Previously Evaluated. 

The proposed performance based 

requirements are an improvement over the 

requirements imposed by the current [TS]. 

Implementation of the proposed SG Program 

will not introduce any adverse changes to the 

plant design basis or postulated accidents 

resulting from potential tube degradation. 

The result of the implementation of the SG 

Program will be an enhancement of SG tube 

performance. Primary to secondary leakage 

that may be experienced during all plant 

conditions will be monitored to ensure it 

remains within current accident analysis 

assumptions. 

The proposed change does not affect the 

design of the SGs, their method of operation, 

or primary or secondary coolant chemistry 

controls. In addition, the proposed change 

does not impact any other plant system or 

component. The change enhances SG 

inspection requirements. 

Therefore, the proposed change does not 

create the possibility of a new or different 

type of accident from any accident 

previously evaluated. 

3. The Proposed Change Does Not Involve 

a Significant Reduction in the Margin of 

Safety. 

The SG tubes in pressurized water reactors 

are an integral part of the reactor coolant 

pressure boundary and, as such, are relied 

upon to maintain the primary system’s 

pressure and inventory. As part of the reactor 

coolant pressure boundary, the SG tubes are 

unique in that they are also relied upon as 

a heat transfer surface between the primary 

and secondary systems such that residual 

heat can be removed from the primary 

system. In addition, the SG tubes isolate the 

radioactive fission products in the primary 

coolant from the secondary system. In 

summary, the safety function of an SG is 

maintained by ensuring the integrity of its 

tubes. 

[SG] tube integrity is a function of the 

design, environment, and the physical 

condition of the tube. The proposed change 

does not affect tube design or operating 

environment. The proposed change is 

expected to result in an improvement in the 

tube integrity by implementing the SG 

Program to manage SG tube inspection, 

assessment, repair, and plugging. The 

requirements established by the SG Program 

are consistent with those in the applicable 

design codes and standards and are an 

improvement over the requirements in the 

current TSs. 

For the above reasons, the margin of safety 

is not changed and overall plant safety will 

be enhanced by the proposed change to the 

TS. 

The NRC staff has reviewed the 

licensee’s incorporation of the above 

analysis by reference and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Item 2: 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. Involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The proposed changes involve adding a 

new definition for RCS [reactor coolant 

system] leakage and rewording certain [TSs] 

for consistency with NUREG–1431, Revision 

3. These changes do not involve any physical 

plant modifications or changes in plant 

operation; consequently, no technical 

changes are being made to the existing TS. 

As such, these changes are administrative in 

nature and do not affect initiators of analyzed 

events or assumed mitigation of accident or 

transient events. Therefore, these changes do 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

The proposed changes involve adding a 

new definition for RCS leakage and 

rewording certain [TSs] for consistency with 

NUREG–1431, Revision 3. These 

administrative changes do not involve 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or changes in methods governing normal 

plant operation. The changes will not impose 

any new or different requirements or 

eliminate any existing requirements. 

Therefore, these changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

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Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 46943 

3. Involve a significant reduction in a 

margin of safety. 

The proposed changes involve adding a 

new definition for RCS leakage and 

rewording certain [TS] for consistency with 

NUREG–1431, Revision 3. The changes are 

administrative in nature and will not involve 

any technical changes. The changes will not 

reduce a margin of safety because they have 

no impact on any safety analysis 

assumptions. Also, since these changes are 

administrative in nature, no question of 

safety is involved. Therefore, the 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 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Lillian M. 

Cuoco, Esq., Senior Counsel, Dominion 

Resources Services, Inc., Millstone 

Power Station, Building 475, 5th Floor, 

Rope Ferry Road, Rt. 156, Waterford, 

Connecticut 06385. 

NRC Branch Chief: Evangelos C. 

Marinos. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (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. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units Nos. 1, 2, and 

3, Maricopa County, Arizona 

Date of application for amendments: 

May 26, 2005, as supplemented by 

letters dated May 23 and June 20, 2006. 

Brief description of amendments: The 

amendments revised Technical 

Specification (TS) 1.1, ‘‘Definitions,’’ TS 

3.4.14, ‘‘RCS [reactor coolant system] 

Operational Leakage,’’ TS 5.5.9, ‘‘Steam 

Generator (SG) Program,’’ and TS 5.6.8, 

‘‘Steam Generator Tube Inspection 

Report,’’ and added a new specification, 

TS 3.4.18, ‘‘Steam Generator (SG) Tube 

Integrity.’’ The changes are consistent 

with TS Task Force (TSTF) Change 

TSTF–449, Revision 4, ‘‘Steam 

Generator Tube Integrity.’’

Date of issuance: July 27, 2006. 

Effective date: As of the date of 

issuance to be implemented within 150 

days from the date of issuance. 

Amendment Nos.: Unit 1–161, Unit 

2–161, Unit 3–161. 

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The 

amendments revised the Operating 

Licenses and the Technical 

Specifications for all three units. 

Date of initial notice in Federal 

Register: July 5, 2005 (70 FR 38714). 

The May 23 and June 20, 2006, 

supplemental letters provided 

additional clarifying information, did 

not expand the scope of the application 

as originally noticed, and did not 

change the NRC staff’s original proposed 

no significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 27, 2006. 

No significant hazards consideration 

comments received: No. 

Carolina Power & Light Company, 

Docket No. 50–261, H. B. Robinson 

Steam Electric Plant, Unit No. 2 

(HBRSEP2), Darlington County, South 

Carolina 

Date of application for amendment: 

January 21, 2005, as supplemented by 

letters dated May 26, 2005, September 

19, 2005, and March 31, 2006. 

Brief description of amendment: The 

amendment approves the 

implementation of the alternative source 

term methodology for a loss-of-coolant 

accident at HBRSEP2. 

Date of issuance: July 11, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No. 207. 

Renewed Facility Operating License 

No. DPR–23. Amendment does not 

revise the Technical Specifications. 

Date of initial notice in Federal 

Register: May 24, 2005 (70 FR 29786). 

The supplemental letters dated May 26, 

2005, September 19, 2005, and March 

31, 2006, provided clarifying 

information that did not change the 

initial proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 11, 2006. 

No significant hazards consideration 

comments received: No. 

Dominion Energy Kewaunee, Inc. Docket 

No. 50–305, Kewaunee Power Station, 

Kewaunee County, Wisconsin 

Date of application for amendment: 

January 12, 2006, as supplemented by 

letter dated June 2, 2006. 

Brief description of amendment: The 

amendment revises the existing steam 

generator (SG) tube surveillance 

program to be consistent with TS Task 

Force (TSTF) Change TSTF–449, 

Revision 4, ‘‘Steam Generator Tube 

Integrity,’’ and the model safety 

evaluation prepared by the Nuclear 

Regulatory Commission (NRC) and 

published in the Federal Register on 

March 2, 2005 (70 FR 10298) under the 

consolidated line item improvement 

process (CLIIP). 

Date of issuance: July 18, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment No.: 188. 

Facility Operating License No. DPR–

43: Amendment revised the Facility 

Operating License and Technical 

Specifications. 

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46944 Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 

Date of initial notice in Federal 

Register: February 14, 2006 (71 FR 

7806). The supplement letter contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination, and did 

not expand the scope of the orginal 

Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 18, 2006. 

No significant hazards consideration 

comments received: No. 

Duke Power Company LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of application for amendments: 

October 27, 2004. 

Brief description of amendments: The 

amendments revised the facility 

operating licenses by removal of license 

condition 2.F, ‘‘Reporting 

Requirements’’, with regard to 

maximum power level, Updated Final 

Safety Analysis Report, antitrust 

conditions, fire protection, and 

additional conditions. 

Date of issuance: July 31, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 230, 226. 

Renewed Facility Operating License 

Nos. NPF–35 and NPF–52: Amendments 

revised the licenses. 

Date of initial notice in Federal 

Register: July 5, 2005 (70 FR 38717). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated July 31, 2006. 

No significant hazards consideration 

comments received: No. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of application for amendment: 

April 17, 2006. 

Brief description of amendment: The 

amendment allows a delay time for 

entering a supported system Technical 

Specification (TS) when the 

inoperability is due solely to an 

inoperable snubber, if risk is assessed 

and managed consistent with the 

program in place for complying with the 

requirements of 10 CFR 50.65(a)(4). 

Limiting Condition for Operation (LCO) 

3.0.8 is added to the TS to provide this 

allowance and define the requirements 

and limitations for its use. 

This change was proposed by the 

industry’s Technical Specification Task 

Force (TSTF) and is designated TSTF–

372, Revision 4. The NRC staff issued a 

notice of opportunity for comment in 

the Federal Register on November 24, 

2004 (69 FR 68412), on possible 

amendments concerning TSTF–372, 

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 May 4, 2005 (70 

FR 23252). The licensee affirmed the 

applicability of the following NSHC 

determination in its application dated 

April 17, 2006. 

Date of issuance: July 11, 2006. 

Effective date: As of its date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment No.: 198. 

Facility Operating License No. NPF–

21: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: May 9, 2006 (71 FR 26998). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 11, 2006. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of application for amendment: 

September 26, 2005, as supplemented 

by letter dated April 11, 2006. 

Brief description of amendment: The 

amendment revises the analysis method 

used for the large-break loss-of-coolant 

accident. 

Date of issuance: July 24, 2006. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 248. 

Facility Operating License No. DPR–

26: The amendment revised the 

Technical Specifications and License. 

Date of initial notice in Federal 

Register: November 8, 2005 (70 FR 

67747). The April 11, 2006, supplement 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the NRC 

staff’s original proposed no significant 

hazards consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 24, 2006. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant (JAFNPP), Oswego 

County, New York 

Date of application for amendment: 

January 26, 2006, as supplemented by 

letter dated April 12, 2006. 

Brief description of amendment: The 

amendment approves the 

implementation of the Boiling Water 

Reactor Vessel and Internals Project 

reactor pressure vessel integrated 

surveillance program as the basis for 

demonstrating the compliance of 

JAFNPP with the requirements of 

Appendix H to Title 10 of the Code of 

Federal Regulations part 50. 

Date of issuance: July 26, 2006. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 285. 

Facility Operating License No. DPR–

59: The amendment revised the 

Updated Final Safety Analysis Report 

and the License. 

Date of initial notice in Federal 

Register: March 14, 2006 (71 FR 

13174). The April 12, 2006, supplement 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the NRC 

staff’s original proposed no significant 

hazards consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 26, 2006. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2, 

Pope County, Arkansas 

Date of application for amendment: 

September 19, 2005. 

Brief description of amendment: The 

amendment modified ANO–2 

Surveillance Requirement TS 3.1.1.4, 

‘‘Moderator Temperature Coefficient,’’

and allowed the use of WCAP–16011–

P–A, ‘‘Startup Test Activity Reduction 

Program.’’

Date of issuance: August 2, 2006. 

Effective date: As of the date of 

issuance to be implemented within 30 

days from the date of issuance. 

Amendment No.: 265. 

Renewed Facility Operating License 

No. NPF–6: Amendment revised the 

Technical Specifications/license. 

Date of initial notice in Federal 

Register: December 6, 2005 (70 FR 

72671). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated August 2, 2006. 

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Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 46945 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2, 

Pope County, Arkansas 

Date of application for amendment: 

September 19, 2005, as supplemented 

by letters dated May 11 and June 19, 

2006. 

Brief description of amendment: The 

amendment revised the existing steam 

generator tube surveillance program to 

be consistent with the U.S. Nuclear 

Regulatory Commission’s approved 

Technical Specification Task Force 

Standard Technical Specification 

Change Traveler, TSTF–449, ‘‘Steam 

Generator Tube Integrity,’’ Revision 4. 

TSTF–449 is part of the consolidated 

line item improvement process. 

Date of issuance: August 2, 2006. 

Effective date: As of the date of 

issuance to be implemented within 90 

days from the date of issuance. 

Amendment No.: 266. 

Renewed Facility Operating License 

No. NPF–6: Amendment revised the 

Technical Specifications and Renewed 

Facility Operating License. 

Date of initial notice in Federal 

Register: January 3, 2006 (71 FR 147). 

The supplements dated May 11 and 

June 19, 2006, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated August 2, 2006. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois 

Date of application for amendments: 

January 25, 2006, as supplemented by 

letter dated May 17, 2006. 

Brief description of amendments: The 

amendment revised the Quad Cities 

licensing basis, as described in the 

Updated Final Safety Analysis Report, 

to allow the use of automatic load tap 

changers to operate in automatic mode 

on the reserve auxiliary transformers to 

compensate for potential offsite power 

voltage fluctuations, in order to ensure 

that acceptable voltage is maintained for 

safety-related equipment. 

Date of issuance: July 24, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 232 and 228. 

Renewed Facility Operating License 

Nos. DPR–29 and DPR–30: The 

amendments revised the License. 

Date of initial notice in Federal 

Register: May 23, 2006 (71 FR 29678). 

The May 17, 2006, supplement 

contained clarifying information and 

did not change the NRC staff’s initial 

proposed finding of no significant 

hazards consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated July 24, 2006. 

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: 

August 23, 2005, as supplemented on 

April 6, 2006. 

Brief description of amendments: The 

amendments extended the licensed lives 

of the Diablo Canyon Power Plant, Unit 

Nos. 1 and 2 reactors by the amount of 

time the licensee had expended to 

perform low-power testing of the 

reactors prior to initial startup. 

Date of issuance: July 17, 2006. 

Effective date: As of its date of 

issuance and shall be implemented 

within 90 days of issuance. 

Amendment Nos.: Unit 1–188; Unit 

2–190. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Facility Operating Licenses. 

Date of initial notice in Federal 

Register: October 11, 2005 (70 FR 

59087). The April 6, 2006, supplemental 

letter provided additional information 

that clarified the application, and did 

not expand the scope of the application 

as originally noticed. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated July 17, 2006. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of application for amendment: 

August 4, 2005, as supplemented by 

letters dated February 9, July 18, and 

August 1, 2006. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 3.7.1.3, ‘‘Ultimate 

Heat Sink,’’ to permit continued plant 

operation if the temperature of the 

ultimate heat sink (UHS) exceeds 89 °F, 

provided the UHS temperature averaged 

over the previous 24-hour period is 

verified at least once per hour to be less 

than or equal to 89 °F, and the UHS 

temperature does not exceed a 

maximum value of 91.4 °F. 

Date of issuance: August 1, 2006. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 168. 

Facility Operating License No. NPF–

57: This amendment revised the TSs. 

Date of initial notice in Federal 

Register: August 30, 2005 (70 FR 

51382). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated August 1, 2006. 

No significant hazards consideration 

comments received: No. 

R.E. Ginna Nuclear Power Plant, LLC, 

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

Power Plant, Wayne County, New York 

Date of application for amendment: 

November 7, 2005, as supplemented on 

May 5, 2006. 

Brief description of amendment: The 

amendment revises Technical 

Specification 3.9.3, ‘‘Containment 

Penetrations,’’ to allow an emergency 

egress door, access door, or roll up door, 

as associated with the equipment hatch 

penetration, to be open, but capable of 

being closed, during core alterations or 

movement of irradiated fuel within 

containment. 

Date of issuance: July 26, 2006. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment No.: 98. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: January 3, 2006 (71 FR 154). 

The May 5, 2006, letter provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 26, 2006. 

No significant hazards consideration 

comments received: No. 

R.E. Ginna Nuclear Power Plant, LLC, 

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

Power Plant, Wayne County, New York 

Date of application for amendment: 

November 18, 2005. 

Brief description of amendment: The 

amendment revises the frequency in 

Technical Specification Surveillance 

Requirement 3.6.6.15, which verifies 

mstockstill on PROD1PC61 with NOTICES

46946 Federal Register / Vol. 71, No. 157 / Tuesday, August 15, 2006 / Notices 

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

2 17 CFR 240.19b–4. 

that each containment spray nozzle is 

unobstructed. The frequency is changed 

from ‘‘10 years’’ to ‘‘following 

maintenance which could result in 

nozzle blockage.’’

Date of issuance: July 31, 2006. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment No.: 99. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

Technical Specifications and the 

License. 

Date of initial notice in Federal 

Register: January 3, 2006 (71 FR 154). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 31, 2006. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50–259 Browns Ferry Nuclear Plant, 

Unit 1, Limestone County, Alabama 

Date of application for amendment: 

December 6, 2004 (TS 428) as 

supplemented by letter dated June 16, 

2005. 

Brief description of amendment: The 

amendment revised the reactor vessel 

Pressure-Temperature curves depicted 

in the Technical Specification (TS) 

Figure 3.4.9–1 and adds a new TS 

Figure 3.4.9–2. 

Date of issuance: July 26, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No.: 256. 

Facility Operating License No. DPR–

33: Amendment revised the TS. 

Date of initial notice in Federal 

Register: January 18, 2005 (70 FR 

2899). The supplement dated June 16, 

2005, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 26, 2006. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50–390, Watts Bar Nuclear Plant, Unit 1, 

Rhea County, Tennessee 

Date of application for amendment: 

December 15, 2005 (TS–05–09), as 

supplemented by letter dated June 7, 

2006. 

Brief description of amendment: The 

amendment revises the Watts Bar 

Nuclear Plant (WBN) Technical 

Specification Surveillance 

Requirements to increase the minimum 

required average ice basket weight, thus, 

increasing the corresponding total 

weight of the stored ice in the WBN ice 

condenser. The changes to the ice basket 

and total ice weights are due to the 

additional energy associated with the 

Replacement Steam Generators. 

Date of issuance: July 25, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to Mode 4 at startup to begin Cycle 

8 fuel cycle. 

Amendment No. 62. 

Facility Operating License No. NPF–

90: Amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: February 14, 2006 (71 FR 

7814). The supplemental letter provided 

clarifying information that was within 

the scope of the initial notice and did 

not change the initial proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 25, 2006. 

No significant hazards consideration 

comments received: No. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application for amendment: 

March 28, 2006. 

Brief description of amendment: The 

amendment revised Technical 

Specification 5.0, ‘‘Administrative 

Controls,’’ by changing a position title 

and department name. 

Date of issuance: July 11, 2006. 

Effective date: As of its date of 

issuance, and shall be implemented 

within 90 days of the date of issuance. 

Amendment No.: 173. 

Facility Operating License No. NPF–

30: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: May 9, 2006 (71 FR 27005). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 11, 2006. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Units 1 and 2, 

Louisa County, Virginia 

Date of application for amendment: 

July 5, 2005, as supplemented by letters 

dated March 30, April 13, and May 11, 

2006. 

Brief description of amendment: The 

amendments revised the Technical 

Specifications (TSs) to add a reference 

in TS 5.65.b, ‘‘Core Operating Limits 

Report (COLR),’’ to permit the use of an 

alternate methodology to perform a 

thermal-hydraulic analysis to predict 

the critical heat flux and departure from 

nucleate boiling ratio for the AREVA 

Advanced Mark-BW fuel in the North 

Anna 1 and 2 cores. 

Date of issuance: July 21, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: 247, 227. 

Renewed Facility Operating License 

Nos. NPF–4 and NPF–7: Amendments 

changed the Licenses and the TSs. 

Date of initial notice in Federal 

Register: August 16, 2005 (70 FR 

48208). The supplements dated March 

30, April 13, and May 11, 2006, 

contained clarifying information only 

and did not change the initial no 

significant hazards consideration 

determination or expand the scope of 

the initial application. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated July 21, 2006. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 8th day 

of August, 2006. 

For the Nuclear Regulatory Commission. 

Catherine Haney, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 06–6921 Filed 8–14–06; 8:45 am] 

BILLING CODE 7590–01–P 

SECURITIES AND EXCHANGE 

COMMISSION 

[Release No. 34–54296; File No. SR–ISE–

2006–30] 

Self-Regulatory Organizations; 

International Securities Exchange, Inc.; 

Order Approving a Proposed Rule 

Change, and Amendment No. 1 

Thereto, Increasing the Linkage 

Inbound Principal Order Fee 

August 9, 2006. 

On June 5, 2006, the International 

Securities Exchange, Inc. (‘‘ISE’’ or 

‘‘Exchange’’) filed with the Securities 

and Exchange Commission 

(‘‘Commission’’), pursuant to Section 

19(b)(1) of the Securities Exchange Act 

of 1934 (‘‘Act’’) 1 and Rule 19b–4 

thereunder,2 a proposed rule change to 

amend its Schedule of Fees in the 

manner described below. On June 29, 

mstockstill on PROD1PC61 with NOTICES