Federal Register / Vol. 72, No. 98 / Tuesday, May 22, 2007 / Notices 28717 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g. 

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

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: May 17, 2007. 

R. Michelle Schroll, 

Office of the Secretary. 

[FR Doc. 07–2559 Filed 5–18–07; 11:28 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from April 27, 

2007, to May 10, 2007. The last 

biweekly notice was published on May 

8, 2007 (72 FR 26173). 

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, Rulemaking, 

Directives and Editing Branch, Division 

of Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be delivered to Room 6D22, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

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

Copies of written comments received 

may be examined at the Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. The filing of 

requests for a hearing and petitions for 

leave to intervene is discussed below. 

Within 60 days after the date of 

publication of this notice, the licensee 

may file a request for a hearing with 

respect to issuance of the amendment to 

the subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

intervene. Requests for a hearing and a 

petition for leave to intervene shall be 

filed in accordance with the 

Commission’s ‘‘Rules of Practice for 

Domestic Licensing Proceedings’’ in 10 

CFR Part 2. Interested persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 


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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 

e-mail 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, Unit 

No. 1, DeWitt County, Illinois 

Date of amendment request: 

December 12, 2006. 

Description of amendment request: 

The proposed amendment would revise 

Surveillance Requirement (SR) 3.3.1.1.8 

and SR 3.3.1.3.2 to increase the interval 

between local power range monitor 

(LPRM) calibrations from 1000 

megawatt-days per ton (MWD/T) 

average core exposure to 2000 MWD/T 

average core exposure. The proposed 

increase in the interval between 

required LPRM calibrations is 

acceptable due to improvements in fuel 

analytical bases, core monitoring 

processes, and nuclear instrumentation. 

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 

surveillance interval for the LPRM 

calibration from 1000 MWD/T average core 

exposure to 2000 MWD/T average core 

exposure. Increasing the frequency interval 

between required LPRM calibrations is 

acceptable due to improvements in fuel 

analytical bases, core monitoring processes, 

and nuclear instrumentation. Therefore, the 

revised surveillance interval continues to 

ensure that the LPRM detector signal will 

continue to be adequately calibrated. 

This change will not alter the operation of 

process variables, structures, systems, or 

components as described in the CPS [Clinton 

Power Station] Updated Safety Analysis 

Report (USAR). The proposed change does 

not alter the initiation conditions or 

operational parameters for the LPRM 

subsystem and there is no new equipment 

introduced by the extension of the LPRM 

calibration interval. The performance of the 

Average Power Range Monitor (APRM) 

system, Oscillation Power Range Monitor 

(OPRM) system, Rod Control and Information 

System (RC&IS) and 3D MONICORE core 

monitoring system is not significantly 

affected by the proposed surveillance interval 

increase. The proposed LPRM calibration 

interval extension will have no significant 

effect on the Reactor Protection System (RPS) 

instrumentation accuracy during power 

maneuvers or transients and will therefore 

not significantly affect the performance of the 

RPS. As such, the probability of occurrence 

for a previously evaluated accident is not 

increased. 

The radiological consequences of an 

accident can be affected by the thermal limits 

existing at the time of the postulated 

accident; however, LPRM chamber exposure 

has no significant affect on the calculated 

thermal limits since LPRM accuracy does not 


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significantly deviate with exposure. For the 

LPRM extended calibration interval, the total 

nodal power uncertainty remains less than 

the uncertainty assumed in the General 

Electric BWR [boiling water reactor] Thermal 

Analysis Basis (GETAB) safety limit, 

maintaining the accuracy of the thermal limit 

calculation. Therefore, the thermal limit 

calculation is not significantly affected by 

LPRM calibration frequency, and 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. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

The performance of the APRM, OPRM, 

RC&IS and 3D MONICORE systems is not 

significantly affected by the proposed LPRM 

surveillance interval increase. The proposed 

change does not affect the control parameters 

governing unit operation or the response of 

plant equipment to transient conditions. The 

proposed amendment does not change or 

introduce any new equipment, modes of 

system operation or failure mechanisms. 

Therefore, 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. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change has no impact on 

equipment design or fundamental operation, 

and there are no changes being made to 

safety limits or safety system allowable 

values that would adversely affect plant 

safety as a result of the proposed LPRM 

surveillance interval increase. The 

performance of the APRM, OPRM, RC&IS and 

3D MONICORE systems is not significantly 

affected by the proposed change. The 

proposed LPRM calibration interval 

extension will have no significant effect on 

RPS instrumentation accuracy during power 

maneuvers or transients and will therefore 

not significantly affect the performance of the 

RPS. The margin of safety can be affected by 

the thermal limits existing at the time of the 

postulated accident; however, uncertainties 

associated with LPRM chamber exposure 

have no significant effect on the calculated 

thermal limits. The thermal limit calculation 

is not significantly affected since LPRM 

sensitivity with exposure is well defined. 

LPRM accuracy remains within the total 

nodal power uncertainty assumed in the 

GETAB, therefore maintaining thermal limits 

and the safety margin. The proposed change 

does not affect safety analysis assumptions or 

initial conditions and therefore, the margin of 

safety in the original safety analyses is 

maintained. 

Based on the above information, 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: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Russell Gibbs. 

AmerGen Energy Company, LLC, Docket 

No. 50–461, Clinton Power Station, Unit 

No. 1, DeWitt County, Illinois 

Date of amendment request: January 

26, 2007. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 3.3.1.1, 

‘‘Reactor Protection System (RPS) 

Instrumentation,’’ Table 3.3.1.1–1, 

‘‘Reactor Protection System 

Instrumentation,’’ Function 8, ‘‘Scram 

Discharge Volume Water Level—High,’’

item b, ‘‘Float Switches,’’ by replacing 

Surveillance Requirement (SR) 3.3.1.1.9 

with SR 3.3.1.1.12. This change will 

effectively revise the surveillance 

frequency for the scram discharge 

volume (SDV) level float switch from 

every 92 days to every 24 months. 

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 TS change involves a change 

in the surveillance frequency for the SDV 

water level float switch channel functional 

test. The proposed TS change does not 

physically impact the plant. The proposed 

change does not affect the design of the SDV 

water level instruments, the operational 

characteristics or function of the instruments, 

the interfaces between the instruments and 

the RPS, or the reliability of the SDV water 

level instruments. The proposed TS change 

does not degrade the performance of, or 

increase the challenges to, any safety systems 

assumed to function in the accident analysis. 

As noted in the Bases to TS 3.3.1.1, even 

though the two types of SDV Water Level—

High Functions are an input to the RPS logic, 

no credit is taken for a scram initiated from 

these functions for any of the design basis 

accidents or transients evaluated in the CPS 

[Clinton Power Station] Updated Safety 

Analysis Report (USAR). An inoperable SDV 

water level instrument is not considered as 

an initiator of any analyzed event. The 

proposed TS change does not impact the 

usefulness of the SRs in evaluating the 

operability of required systems and 

components, or the way in which the 

surveillances are performed. In addition, the 

frequency of surveillance testing is not 

considered an initiator of any analyzed 

accident, nor does a revision to the frequency 

introduce any accident initiators. Therefore, 

the proposed change does not involve a 

significant increase in the 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 consequences of a previously 

evaluated accident are not significantly 

increased by the proposed change. The 

proposed change does not affect the 

performance of any equipment credited to 

mitigate the radiological consequences of an 

accident. The risk assessment of the 

proposed changes has concluded that there is 

an insignificant increase in the core damage 

frequency as well as the total population 

dose rate. Historical review of surveillance 

test results and associated maintenance 

records did not find evidence of failures that 

would invalidate the above conclusions. 

Therefore, the proposed change does not 

alter the ability to detect and mitigate events 

and, as such, does not involve a significant 

increase in the consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

The proposed TS change does not 

introduce any failure mechanisms of a 

different type than those previously 

evaluated, since there are no physical 

changes being made to the facility. No new 

or different equipment is being installed. No 

installed equipment is being operated in a 

different manner. There is no change being 

made to the parameters within which CPS is 

operated. There are no setpoints at which 

protective or mitigative actions are initiated 

that are affected by this proposed action. The 

change does not alter assumptions made in 

the safety analysis. This proposed action will 

not alter the manner in which equipment 

operation is initiated, nor will the function 

demands on credited equipment be changed. 

No alteration in the procedures, which 

ensure the unit remains within analyzed 

limits, is proposed, and no change is being 

made to procedures relied upon to respond 

to an off-normal event. As a result, no new 

failure modes are being introduced. The way 

surveillance tests are performed remains 

unchanged. A historical review of 

surveillance test results and associated 

maintenance records indicated there was no 

evidence of any failures that would 

invalidate the above conclusions. 

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. 

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 


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proposed TS change involves a change in the 

surveillance frequency for the SDV water 

level float switch channel functional test. 

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. The 

proposed change does not significantly 

impact the condition or performance of 

structures, systems, and components relied 

upon for accident mitigation. The proposed 

change does not result in any hardware 

changes or in any changes to the analytical 

limits assumed in accident analyses. Existing 

operating margin between plant conditions 

and actual plant setpoints is not significantly 

reduced due to these changes. The proposed 

change does not significantly impact any 

safety analysis assumptions or results. 

AmerGen has conducted a risk assessment 

to determine the impact of a change to the 

SDV water level instrument surveillance 

frequency from the current once every 92 

days to once every 24 months for the risk 

measures of Core Damage Frequency (CDF) 

and Large Early Release Frequency (LERF). 

This assessment indicated that the proposed 

CPS surveillance frequency extension has a 

very small change in risk to the public and 

is an acceptable plant change from a risk 

perspective. 

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: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Russell Gibbs. 

Carolina Power & Light Company, et al., 

Docket No. 50–400, Shearon Harris 

Nuclear Power Plant, Unit 1, Wake and 

Chatham Counties, North Carolina 

Date of amendment request: April 30, 

2007. 

Description of amendment request: 

The amendment will revise the 

technical specifications to use other 

narrow range containment sump water 

level instrumentation rather than the 

existing redundant instruments to allow 

installation of new emergency core 

cooling system recirculation sumps 

strainers as specified in the Nuclear 

Regulatory Commission Generic Letter 

2004–02, Potential Impact of Debris 

Blockage on Emergency Recirculation 

during Design Basis Accidents at 

Pressurized Water Reactors. 

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 amendment does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated, and it does not change an accident 

previously evaluated in the Final Safety 

Analysis Report (FSAR). The use of other 

narrow range containment sump water level 

instruments rather than the existing narrow 

range containment recirculation sump water 

level instruments, which have level elements 

located inside the emergency core cooling 

system (ECCS) recirculation sumps, will 

continue to ensure that acceptable narrow 

range containment sump water level 

monitoring is maintained during postaccident conditions. Operation of the 

containment spray and residual heat removal 

systems is unchanged as a result of the 

proposed amendment. The level elements 

associated with the existing narrow range 

containment recirculation sump water level 

instruments are not accident initiators, and 

the FSAR does not credit these level 

elements in the dose analyses for loss-ofcoolant accidents. The proposed amendment 

does not adversely affect the ability of 

structures, systems, or components (SSCs) to 

perform their design function. SSCs required 

for post-accident recirculation remain 

capable of performing their design functions. 

Therefore, this amendment 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 amendment does not create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated, and it does not change an accident 

previously evaluated in the Final Safety 

Analysis Report (FSAR). The use of other 

narrow range containment sump water level 

instruments rather than the existing narrow 

range containment recirculation sump water 

level instruments supports the replacement 

of the existing containment recirculation 

sump screens with new strainers in 

accordance with the response to Generic 

Letter 2004–02, Potential Impact of Debris 

Blockage on Emergency Recirculation during 

Design Basis Accidents at Pressurized-Water 

Reactors. The proposed amendment does not 

change the design function or the operation 

of the containment spray and residual heat 

removal systems associated with the 

containment recirculation sumps. The 

proposed amendment does not create new 

failure mechanisms or malfunctions or 

accident initiators. The proposed amendment 

will continue to ensure that acceptable 

narrow range containment sump water level 

monitoring is maintained during postaccident conditions, and that SSCs required 

for post-accident recirculation remain 

capable of performing their design functions. 

Therefore, this amendment 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 amendment does not involve 

a significant reduction in a margin of safety. 

The proposed amendment does not adversely 

affect a plant safety limit or a limiting safety 

system setting, and does not alter a design 

basis limit for a parameter evaluated in the 

FSAR. The use of other narrow range 

containment sump water level instruments, 

which meet the requirements of the FSAR, 

rather than the existing narrow range 

containment recirculation sump water level 

instruments, will continue to ensure that 

acceptable narrow range containment sump 

water level monitoring is maintained during 

post-accident conditions. The proposed 

amendment does not adversely affect the 

ability of SSCs to perform their design 

functions or the reliability of equipment to 

mitigate accidents evaluated in the FSAR. 

The proposed amendment will continue to 

ensure that SSCs required for post-accident 

recirculation remain capable of performing 

their design functions. 

Therefore, this 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: David T. 

Conley, Associate General Counsel II—

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Branch Chief: Thomas H. Boyce. 

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: April 24, 

2007. 

Description of amendment request: 

The proposed change will add 

Optimized ZIRLOTM as an acceptable 

fuel rod cladding material in the 

Waterford Steam Electric Station, Unit 3 

(Waterford 3), Technical Specification 

(TS) 5.3.1, ‘‘Fuel Assemblies.’’ TS 5.3.1 

currently identifies, in part, Zircaloy or 

ZIRLOTM fuel rod cladding as the 

allowable fuel rod cladding material. 

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 


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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 NRC-approved topical report WCAP–

12610–P–A and CENPD–404–P–A, 

Addendum 1–A, ‘‘Optimized ZIRLOTM,’’

prepared by Westinghouse Electric Company, 

LLC (Westinghouse), addresses Optimized 

ZIRLOTM and demonstrates that Optimized 

ZIRLOTM has essentially the same properties 

as currently licensed ZIRLOTM. The fuel 

cladding itself is not an accident initiator and 

does not affect accident probability. Use of 

Optimized ZIRLOTM fuel cladding has been 

shown to meet all 10 CFR 50.46 design 

criteria and, therefore, will not increase the 

consequences of an accident. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Use of Optimized ZIRLOTM clad fuel will 

not result in changes in the operation or 

configuration of the facility. Topical report 

WCAP–12610–P–A and CENPD–404–P–A 

demonstrated that the material properties of 

Optimized ZIRLOTM are similar to those of 

standard ZIRLOTM. Therefore, Optimized 

ZIRLOTM fuel rod cladding will perform 

similarly to those fabricated from standard 

ZIRLOTM, thus precluding the possibility of 

the fuel becoming an accident initiator and 

causing a new or different type of accident. 

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 will not involve a 

significant reduction in the margin of safety 

because it has been demonstrated that the 

material properties of the Optimized 

ZIRLOTM are not significantly different from 

those of standard ZIRLOTM. Optimized 

ZIRLOTM is expected to perform similarly to 

standard ZIRLOTM for all normal operating 

and accident scenarios, including both lossof-coolant accident (LOCA) and non-LOCA 

scenarios. For LOCA scenarios, where the 

slight difference in Optimized ZIRLOTM

material properties relative to standard 

ZIRLOTM could have some impact on the 

overall accident scenario, plant-specific 

LOCA analyses using Optimized ZIRLOTM

properties will be performed prior to the use 

of fuel assemblies with fuel rods containing 

Optimized ZIRLOTM. These LOCA analyses 

will demonstrate that the acceptance criteria 

of 10 CFR 50.46 will be satisfied when 

Optimized ZIRLOTM fuel rod cladding is 

implemented. 

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: Terence A. 

Burke, Associate General Council—

Nuclear Entergy Services, Inc., 1340 

Echelon Parkway, Jackson, Mississippi 

39213. 

NRC Branch Chief: Thomas G. Hiltz. 

Nine Mile Point Nuclear Station 

(NMPNS), LLC, Docket No. 50–410, Nine 

Mile Point Nuclear Station Unit No. 2 

(NMP2), Oswego County, New York 

Date of amendment request: March 

30, 2007. 

Description of amendment request: 

The proposed amendment would 

change the NMP2 Technical 

Specifications to reflect an expanded 

operating domain resulting from 

implementation of Average Power 

Range Monitor/Rod Block Monitor/ 

Technical Specifications/Maximum 

Extended Load Line Analysis (ARTS/ 

MELLLA). The Average Power Range 

Monitor (APRM) flow-biased simulated 

thermal power Allowable Value would 

be revised to permit operation in the 

MELLLA region. The current flowbiased Rod Block Monitor (RBM) would 

be replaced by a power dependent RBM, 

which also would require new 

Allowable Values. The flow-biased 

APRM simulated thermal power 

setdown requirement would be replaced 

by more direct power and flow 

dependent thermal limits 

administration. The Surveillance 

Requirement for the standby liquid 

control (SLC) system would be revised 

to require each SLC pump to deliver 

required flow at a discharge pressure 

≥1325 psig in lieu of ≥1320 psig; the 

SLC relief valve setpoint would be 

increased from 1394 psig to 1400 psig. 

Finally, the proposed amendment 

employs a new model for performing 

the anticipated transients without scram 

(ATWS) analysis for ARTS/MELLLA 

conditions. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

(1) Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change eliminates the APRM 

flow-biased simulated thermal power 

setdown requirement and substitutes power 

and flow dependent adjustments to the 

Minimum Critical Power Ratio (MCPR) and 

Linear Heat Generation Rate (LHGR) thermal 

limits. Thermal limits will be determined 

using NRC [Nuclear Regulatory Commission] 

approved analytical methods. The proposed 

change will have no effect upon any accident 

initiating mechanism. The power and flow 

dependent adjustments will ensure that the 

MCPR safety limit will not be violated as a 

result of any Anticipated Operational 

Occurrence (AOO), and that the fuel thermal 

and mechanical design bases will be 

maintained. 

The proposed change also expands the 

power and flow operating domain by relaxing 

the restrictions imposed by the formulation 

of the APRM flow-biased simulated thermal 

power Allowable Value and the replacement 

of the current flow-biased RBM with a new 

power dependent RBM. The APRM and RBM 

are not involved in the initiation of any 

accident, and the APRM flow-biased 

simulated thermal power function is not 

credited in any NMP2 safety analyses. The 

proposed change will not introduce any 

initial conditions that would result in NRC 

approved criteria being exceeded and the 

APRM and RBM will remain capable of 

performing their design functions. 

The Standby Liquid Control (SLC) System 

is provided to mitigate anticipated transients 

without scram (ATWS) events and, as such, 

is not considered an initiator of an ATWS 

event or any other analyzed accident. The 

revised SLC discharge pump test pressure 

neither reduces the ability of the SLC system 

to respond to or mitigate an ATWS event nor 

increases the likelihood of a system 

malfunction that could increase the 

consequences of an accident. 

Based on the above discussion, it is 

concluded that 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 eliminates the APRM 

flow-biased simulated thermal power 

setdown requirement and substitutes power 

and flow dependent adjustments to the 

MCPR and LHGR thermal limits. Because the 

thermal limits will continue to be met, no 

analyzed transient event will escalate into a 

new or different type of accident due to the 

initial starting conditions permitted by the 

adjusted thermal limits. 

The proposed change also expands the 

power and flow operating domain by relaxing 

the restrictions imposed by the formulation 

of the APRM flow-biased simulated thermal 

power Allowable Value and the replacement 

of the current flow-biased RBM with a new 

power dependent RBM. Changing the 

formulation for the APRM flow-biased 

simulated thermal power Allowable Value 

and changing from a flow-biased RBM to a 

power dependent RBM does not change their 

respective functions and manner of 


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operation. The change does not introduce a 

sequence of events or introduce a new failure 

mode that would create a new or different 

[kind] of accident. While not credited, the 

APRM flow-biased simulated thermal power 

Allowable Value and associated scram trip 

setpoint will continue to initiate a scram to 

protect the MCPR safety limit. The power 

dependent RBM will prevent rod withdrawal 

when the power dependent RBM rod block 

setpoint is reached. No new failure 

mechanisms, malfunctions, or accident 

initiators are being introduced by the 

proposed change. In addition, operating 

within the expanded power flow map will 

not require any systems, structures or 

components to function differently than 

previously evaluated and will not create 

initial conditions that would result in a new 

or different kind of accident from any 

accident previously evaluated. 

The proposed change to the SLC pump test 

discharge pressure is consistent with the 

functional requirements of the ATWS rule 

(10 CFR 50.62). This proposed change does 

not involve the installation of any new or 

different type of equipment, does not 

introduce any new modes of plant operation, 

and does not change any methods governing 

normal plant operation. 

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 eliminates the APRM 

flow-biased simulated thermal power 

setdown requirement and substitutes power 

and flow dependent adjustments to the 

MCPR and LHGR thermal limits. 

Replacement of the APRM setdown 

requirement with power and flow dependent 

adjustments to the MCPR and LHGR thermal 

limits will continue to ensure that margins to 

the fuel cladding Safety Limit are preserved 

during operation at other than rated 

conditions. Thermal limits will be 

determined using NRC approved analytical 

methods. The power and flow dependent 

adjustments will ensure that the MCPR safety 

limit will not be violated as a result of any 

AOO, and that the fuel thermal and 

mechanical design bases will be maintained. 

The proposed change also expands the 

power and flow operating domain by relaxing 

the restrictions imposed by the formulation 

of the APRM flow-biased simulated thermal 

power Allowable Value and the replacement 

of the current flow-biased RBM with a new 

power dependent RBM. The APRM flowbiased simulated thermal power Allowable 

Value and associated scram trip setpoint will 

continue to initiate a scram to protect the 

MCPR safety limit. The RBM will continue to 

prevent rod withdrawal when the power 

dependent RBM rod block setpoint is 

reached. The MCPR and LHGR thermal limits 

will be developed to ensure that fuel thermal 

mechanical design bases remain within the 

licensing limits during a control rod 

withdrawal error event and to ensure that the 

MCPR safety limit will not be violated as a 

result of a control rod withdrawal error 

event. Operation in the expanded operating 

domain will not alter the manner in which 

safety limits, limiting safety system settings, 

or limiting conditions for operation are 

determined. AOOs and postulated accidents 

within the expanded operating domain will 

continue to be evaluated using NRC 

approved methods. The 10 CFR 50.46 

acceptance criteria for the performance of the 

ECCS [emergency core cooling system] 

following postulated LOCAs [loss-of-coolant 

accidents] will continue to be met. 

The proposed change to the SLC pump 

discharge test pressure does not alter the 

results of any accident analyses. The 

proposed change is consistent with the 

functional requirements of the ATWS rule 

(10 CFR 50.62). The ability of the SLCS to 

respond to and mitigate an ATWS event is 

not affected. 

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: Mark J. 

Wetterhahn, Esquire, Winston & Strawn, 

1700 K Street, NW., Washington, DC 

20006. 

NRC Branch Chief: Mark G. Kowal. 

Southern California Edison Company, et 

al., Docket Nos. 50–361 and 50–362, 

San Onofre Nuclear Generating Station, 

Units 2 and 3, San Diego County, 

California 

Date of amendment requests: April 

17, 2007. 

Description of amendment requests: A 

change is proposed to the standard 

technical specifications (STS) (NUREGs 

1430 through 1434) and plant-specific 

technical specifications (TS), to 

strengthen TS requirements regarding 

control room envelope (CRE) 

habitability by changing the action and 

surveillance requirements associated 

with the limiting condition for 

operation operability requirements for 

the CRE emergency ventilation system, 

and by adding a new TS administrative 

controls program on CRE habitability. 

Accompanying the proposed TS change 

are appropriate conforming technical 

changes to the TS Bases. The proposed 

revision to the Bases also includes 

editorial and administrative changes to 

reflect applicable changes to the 

corresponding STS Bases, which were 

made to improve clarity, conform with 

the latest information and references, 

correct factual errors, and achieve more 

consistency among the STS NUREGs. 

The proposed revision to the TS and 

associated Bases is consistent with STS 

as revised by TS Task Force (TSTF)–

448, Revision 3, ‘‘Control Room 

Envelope Habilitability.’’

The proposed amendment would 

revise the TS Improvement To Modify 

Requirements Regarding CRE 

Habitability using the Consolidated Line 

Item Improvement Process, based on the 

NRC-approved to TSTF–448, Revision 3. 

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on October 17, 2006 (71 FR 

61075), on possible amendments 

adopting TSTF–448, 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 January 17, 2007 (72 FR 

2022). The licensee affirmed the 

applicability of the following NSHC 

determination in its application dated 

April 17, 2007. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below: 

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change does not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, or 

configuration of the facility. The proposed 

change does not alter or prevent the ability 

of structures, systems, and components 

(SSCs) to perform their intended function to 

mitigate the consequences of an initiating 

event within the assumed acceptance limits. 

The proposed change revises the TS for the 

CRE emergency ventilation system, which is 

a mitigation system designed to minimize 

unfiltered air leakage into the CRE and to 

filter the CRE atmosphere to protect the CRE 

occupants in the event of accidents 

previously analyzed. An important part of 

the CRE emergency ventilation system is the 

CRE boundary. The CRE emergency 

ventilation system is not an initiator or 

precursor to any accident previously 

evaluated. Therefore, the probability of any 

accident previously evaluated is not 

increased. Performing tests to verify the 

operability of the CRE boundary and 

implementing a program to assess and 

maintain CRE habitability ensure that the 

CRE emergency ventilation system is capable 

of adequately mitigating radiological 

consequences to CRE occupants during 

accident conditions, and that the CRE 

emergency ventilation system will perform as 

assumed in the consequence analyses of 

design basis accidents. Thus, the 

consequences of any accident previously 

evaluated are not increased. Therefore, the 

proposed change does not involve a 


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significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Accident 

Previously Evaluated 

The proposed change does not impact the 

accident analysis. The proposed change does 

not alter the required mitigation capability of 

the CRE emergency ventilation system, or its 

functioning during accident conditions as 

assumed in the licensing basis analyses of 

design basis accident radiological 

consequences to CRE occupants. No new or 

different accidents result from performing the 

new surveillance or following the new 

program. The proposed change does not 

involve a physical alteration of the plant (i.e., 

no new or different type of equipment will 

be installed) or a significant change in the 

methods governing normal plant operation. 

The proposed change does not alter any 

safety analysis assumptions and is consistent 

with current plant operating practice. 

Therefore, this change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The proposed change does not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The proposed 

change does not affect safety analysis 

acceptance criteria. The proposed change 

will not result in plant operation in a 

configuration outside the design basis for an 

unacceptable period of time without 

compensatory measures. The proposed 

change does not adversely affect systems that 

respond to safely shut down the plant and to 

maintain the plant in a safe shutdown 

condition. 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 requests involve no 

significant hazards consideration. 

Attorney for licensee: Douglas K. 

Porter, Esquire, Southern California 

Edison Company, 2244 Walnut Grove 

Avenue, Rosemead, California 91770. 

NRC Branch Chief: Thomas G. Hiltz. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: February 

28, 2007. 

Brief description of amendments: The 

proposed amendment request would 

revise the language of Technical 

Specification (TS) 3.7.1.2, ‘‘Auxiliary 

Feedwater System,’’ Action b from 

‘‘MODE 3 may be entered with an 

inoperable turbine-driven auxiliary 

feedwater pump for the purposes of 

performing Surveillance Requirement 

4.7.1.2.1a.2’’ to ‘‘MODE 3 may be 

entered with an inoperable turbinedriven auxiliary feedwater pump.’’

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 deletion of the existing 

words in TS 3.7.1.2 Action b is an 

administrative change that will clarify the 

Licensing Basis for the turbine-driven 

auxiliary feedwater pump. Since this change 

does not change the Licensing Basis for TS 

3.7.1.2, this change cannot affect the 

probability or consequence of any accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed deletion of the existing 

words in TS 3.7.1.2 Action b is an 

administrative change that will clarify the 

Licensing Basis for the turbine-driven 

auxiliary feedwater pump. Since this change 

does not change the Licensing Basis for TS 

3.7.1.2, this change cannot affect 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 deletion of the existing 

words in TS 3.7.1.2 Action b is an 

administrative change that will clarify the 

Licensing Basis for the turbine-driven 

auxiliary feedwater pump. Since this change 

does not change the Licensing Basis for TS 

3.7.1.2, this change cannot 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: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Branch Chief: Thomas G. Hiltz. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant, Units 1 and 2, Hamilton 

County, Tennessee 

Date of amendment request: April 5, 

2007. 

Description of amendment request: 

The proposed amendments would 

revise technical specifications (TSs) to 

change the surveillance frequency for 

the turbine trip functions of the reactor 

trip system instrumentation. The 

current frequency is prior to each 

reactor startup and the proposed change 

will revise this to be prior to exceeding 

the Permissive P–9 interlock whenever 

the unit has been in hot standby. The 

proposed change is consistent with 

NRC-approved Technical Specification 

Task Force Traveler TSTF–311, as 

incorporated into the latest revision of 

Standard TSs (NUREG–1431, 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 change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes revise the 

surveillance frequency for reactor trip 

functions from a turbine trip event. These 

changes do not alter these functions 

physically or how they are maintained. 

Delaying the performance of the surveillance 

up to the P–9 interlock will continue to 

ensure operability of the function before the 

plant is in a condition that would benefit 

from the associated actuation. The 

incorporation of a surveillance frequency that 

is consistent with the applicability for the 

function eliminates potential misapplication 

of the TS requirements. The frequency 

changes support turbine trip operability 

during plant startup and are consistent with 

their ability to perform the reactor trip 

functions. Since these changes will not affect 

the ability of these trips to perform the 

initiation of reactor trips when appropriate, 

the off-site dose consequences for an accident 

will not be impacted. Equally, the potential 

to cause an accident is not affected because 

no plant system or component has been 

altered by the proposed changes. Therefore, 

the proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes only affect the 

surveillance frequency requirement for the 

turbine trip functions. This does not affect 


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any physical features of the plant or the 

manner in which these functions are utilized. 

The proposed surveillance frequency will 

require the functions to be verified operable 

before the turbine trip functions are 

applicable and able to perform their trip 

functions. Delaying the performance of the 

surveillance up to the P–9 interlock will 

continue to ensure operability of the function 

before the plant is in a condition that would 

benefit from the associated actuation. 

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 changes do not alter any 

plant setpoints or functions that are assumed 

to actuate in the event of postulated 

accidents. In fact, the proposed changes do 

not alter any plant feature and only alter the 

requirements for when the function must be 

verified to be operable through surveillance 

testing. The proposed changes ensure the 

functionality of the turbine trips when 

assumed in the analysis 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 

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: Thomas H. Boyce. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request: 

December 19, 2006. 

Brief description of amendments: The 

proposed amendment request would 

revise the requirements in Technical 

Specification (TS) 5.5.8, ‘‘Inservice 

Testing Program,’’ to update references 

to the American Society of Mechanical 

Engineers (ASME) Boiler and Pressure 

Vessel Code, Section XI, as the source 

of requirements for the inservice testing 

of ASME Code Class 1, 2, and 3 pumps 

and valves, and address the 

applicability of Surveillance 

Requirement 3.0.2 to other normal and 

accelerated frequencies specified as 2 

years or less in the Inservice Testing 

Program. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed [change] involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes revise TS 5.5.8, 

‘‘Inservice Testing Program,’’ for consistency 

with the requirements of 10 CFR 50.55a(f)(4) 

regarding the inservice testing of pumps and 

valves. 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 changes do not impact any 

accident initiators or analyzed events or 

assumed mitigation of accident or transient 

events. They do not involve the addition or 

removal of any equipment, or any design 

changes to the facility. Therefore, the 

proposed changes do not represent 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 TS 5.5.8, 

‘‘Inservice Testing Program,’’ for consistency 

with the requirements of 10 CFR 50.55a(f) 

regarding the inservice testing of pumps and 

valves. The proposed changes incorporate 

revisions to the ASME Code that result in a 

net improvement in the measures for testing 

pumps and valves. 

The proposed changes do not involve a 

modification to the physical configuration of 

the plant (i.e., no new equipment will be 

installed) or change in the methods 

governing normal plant operation. The 

proposed changes will not impose any new 

or different requirements or introduce a new 

accident initiator, accident precursor, or 

malfunction mechanism. Additionally, there 

is no change in the types or increases in the 

amounts of any effluent that may be released 

off-site and there is no increase in individual 

or cumulative occupational exposure. 

Therefore, these proposed changes do not 

create the possibility of an accident of a 

different kind than previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes revise TS 5.5.8, 

‘‘Inservice Testing Program,’’ for consistency 

with the requirements of 10 CFR 50.55a(f)(4) 

regarding the inservice testing of pumps and 

valves. The proposed changes incorporate 

revisions to the ASME Code that result in a 

net improvement in the measures for testing 

pumps and valves. The safety function of the 

affected pumps and valves will be 

maintained. Therefore, these 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: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Branch Chief: Thomas G. Hiltz. 

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 51.22(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, 


jlentini on PROD1PC65 with NOTICES
Federal Register / Vol. 72, No. 98 / Tuesday, May 22, 2007 / Notices 28725 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

Consumers Energy Company, Entergy 

Nuclear Palisades, LLC, and Entergy 

Nuclear Operations, Inc., Docket No. 

50–155, Big Rock Point Facility, 

Charlevoix County, Michigan 

Date of application for amendment: 

October 31, 2006. 

Brief description of amendment: The 

license amendment reflects the changes 

in ownership and operating authority 

for the Big Rock Facility and its 

Independent Spent Fuel Storage 

Installation. 

Date of issuance: April 11, 2007. 

Effective date: As of the date of 

issuance. 

Amendment No.: 127. 

Facility Operating License No. DPR–

06: The license amendment reflects the 

changes in ownership and operating 

authority for the Big Rock Facility and 

its Independent Spent Fuel Storage 

Installation. 

Date of initial notice in Federal 

Register: January 30, 2007 (72 FR 

4302–4303). The Commission’s related 

evaluation of the amendment is 

contained in a safety evaluation report 

dated April 6, 2007, which is accessible 

to members of the public through 

ADAMS (Accession Number 

ML070920385). 

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 10, 2007, as supplemented by 

letters dated April 5 and 27, 2007. 

Brief description of amendment: The 

amendment modifies the emergency 

diesel generators short-time load testing 

requirements. 

Date of issuance: May 1, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 191. 

Facility Operating License No. DPR–

43: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: February 5, 2007 ( 72 FR 

5303). The supplemental letters 

provided clarifying information that did 

not expand the scope of the original 

application or change the initial 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated May 1, 2007. 

No significant hazards consideration 

comments received: No. 

Duke Power Company LLC, Docket Nos. 

50–369 and 50–370, McGuire Nuclear 

Station, Units 1 and 2, Mecklenburg 

County, North Carolina 

Date of application for amendments: 

April 11, 2006. Brief description of amendments: The 

amendments revised an organizational 

description in the Technical 

Specification Section 5.2.1, ‘‘Onsite and 

Offsite Organizations.’’ The change 

revises the title of Executive Vice 

President to Group Vice President to 

reflect title changes made by the 

licensee following the indirect transfer 

of the facility operating licenses. The 

indirect transfer was reviewed and 

approved by the NRC. This change is 

solely administrative in nature. Date of issuance: April 13, 2007. Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 239, 221. 

Renewed Facility Operating License 

Nos. NPF–9 and NPF–17: Amendments 

revised the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: March 13, 2007 (72 FR 

11387). The Commission’s related 

evaluation, final no significant hazards 

consideration finding, and State 

consultation of the amendments is 

contained in a Safety Evaluation dated 

April 13, 2007. 

No significant hazards consideration 

comments received: No. 

Duke Power Company LLC, Docket Nos. 

50–369 and 50–370, McGuire Nuclear 

Station, Units 1 and 2, Mecklenburg 

County, North Carolina 

Date of application for amendments: 

March 8, 2007, as supplemented March 

27, April 13, and May 3, 2007. 

Brief description of amendments: The 

amendments revise the McGuire 

Nuclear Station, Units 1 and 2, 

Technical Specification 3.5.2.8, and the 

associated Bases and authorize changes 

to the Updated Final Safety Analysis 

Report (USFAR) concerning 

modifications to the emergency core 

cooling system sump. 

Date of issuance: May 4, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 240, 222. 

Renewed Facility Operating License 

Nos. NPF–9 and NPF–17: Amendments 

revised the licenses and the technical 

specifications and authorize changes to 

the UFSAR. 

Date of initial notice in Federal 

Register: March 19, 2007 (72 FR 

12835). 

The supplements dated March 27, 

April 13, and May 3, 2007, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. The 

Commission’s related evaluation, final 

no significant hazards finding, and state 

consultation of the amendments are 

contained in a Safety Evaluation dated 

May 4, 2007. 

No significant hazards consideration 

comments received: No. 

Entergy Gulf States, Inc., and Entergy 

Operations, Inc., Docket No. 50–458, 

River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: 

September 19, 2006, as supplemented 

by letter dated February 28, 2007. 

Brief description of amendment: The 

amendment revised River Bend Station 

(RBS), Unit 1, Technical Specifications 

(TS) Surveillance Requirement (SR) 

3.6.1.3.5 to replace the currently 

specified frequency for leak testing 

containment purge supply and exhaust 

isolation valves with resilient seal 

materials with a requirement to test 

these valves in accordance with the 

RBS’s Primary Containment Leakage 

Rate Testing Program. RBS’s Primary 

Containment Leakage Rate Testing 

Program is implemented in accordance 

with the Title 10 of the Code of Federal 

Regulations, Part 50, Appendix J, 

Option B, and Regulatory Guide (RG) 

1.163, ‘‘Performance-Based Containment 

Leak Test Program,’’ dated September 

1995. RG 1.163 allows a nominal test 

interval of 30 months for containment 

purge and vent valves. 

Date of issuance: May 3, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 60 

days from the date of issuance. 

Amendment No.: 152. 

Facility Operating License No. NPF–

47: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 24, 2006 (71 FR 

62310). The supplement dated February 

28, 2007, 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 May 3, 2007. 


jlentini on PROD1PC65 with NOTICES
28726 Federal Register / Vol. 72, No. 98 / Tuesday, May 22, 2007 / Notices 

No significant hazards consideration 

comments received: No. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of application for amendment: 

May 31, 2005, as supplemented by 

letters dated February 8, 2006, and 

January 5, February 13, February 22, 

and March 22, 2007. 

Brief description of amendment: The 

amendment modifies Technical 

Specification (TS) Sections 3.8.1, ‘‘AC 

[Alternating Current] Sources—

Operating,’’ 3.8.4, ‘‘DC [Direct Current] 

Sources—Operating,’’ 3.8.5, ‘‘DC 

Sources—Shutdown,’’ 3.8.6, ‘‘Battery 

Cell Parameters,’’ and 5.5, ‘‘Programs 

and Manuals.’’ The change incorporates 

clarifying requirements in surveillance 

testing of diesel generators and new 

actions for an inoperable battery 

charger. The change includes a revision 

to the Administrative Program to be 

consistent with Institute of Electrical 

and Electronics Engineers Standard 

450–2002, and changes consistent with 

TS Task Force (TSTF) Traveler TSTF–

360, Revision 1, ‘‘DC Electrical 

Rewrite,’’ and TSTF–283, Revision 3, 

‘‘Modify Section 3.8 Mode Restriction 

Notes.’’

Date of issuance: May 1, 2007. 

Effective date: As of its date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment No.: 204. 

Facility Operating License No. NPF–

21: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: February 27, 2007 (72 FR 

8803). The supplemental letters dated 

February 8, 2006, and January 5, 

February 13, February 22, and March 

22, 2007, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. The 

Commission’s related evaluation of the 

amendment is contained in a Safety 

Evaluation dated May 1, 2007. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station,Plymouth County, 

Massachusetts. 

Date of amendment request: October 

18, 2005, as supplemented by letter 

dated February 23, 2007. 

Description of amendment request: 

The proposed amendment revised 

applicability requirements related to 

single control rod withdrawal 

allowances in shutdown modes. The 

amendment also corrected a 

typographical error and administratively 

relocated the existing TS 3/4.10.D, 

‘‘Multiple Control Rod Removal,’’ to TS 

3/4.14.E to be consistent with the intent 

and presentation of special operations. 

Date of issuance: April 25, 2007. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 90 days. 

Amendment No.: 228. 

Facility Operating License No. DPR–

35: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: January 3, 2006 (71 FR 148). 

The February 23, 2007, supplemental 

letter provided additional information 

that clarified the application, but 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 April 25, 2007. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–254, Quad Cities 

Nuclear Power Station, Unit 1, Rock 

Island County, Illinois 

Date of application for amendments: 

January 16, 2007, as supplemented by 

letter dated April 10, 2007. 

Brief description of amendment: The 

amendment revises the values of the 

safety limit minimum critical power 

ratio (SLMCPR) in the Quad Cities 

Nuclear Power Station (Quad Cities), 

Unit 1, Technical Specification (TS) 

Section 2.1.1, ‘‘Reactor Core SLs [Safety 

Limits].’’ Specifically, the proposed 

change would require that for Unit 1, 

the minimum critical power ratio shall 

be greater than or equal to 1.11 for two 

recirculation loop operation, or greater 

than or equal to 1.13 for single 

recirculation loop operation. This 

change is needed to support the next 

cycle of operation for Quad Cities, Unit 

1. 

Date of issuance: May 2, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to startup from Q1R19 Refueling 

Outage. 

Amendment No.: 234. 

Renewed Facility Operating License 

No. DPR–29: The amendments revised 

the Technical Specifications and 

License. 

Date of initial notice in Federal 

Register: March 13, 2007 (71 FR 

11388). The supplements 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 May 2, 2007. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit No. 1, 

Lake County, Ohio 

Date of application for amendment: 

February 14, 2006, as supplemented by 

letters dated October 17, 2006, and 

February 8, 2007. 

Brief description of amendment: The 

amendment revised Perry Nuclear 

Power Plant, Unit No. 1, Technical 

Specifications (TSs) to change the 

frequency of the Mode 5 Intermediate 

Range Monitoring Instrumentation 

CHANNEL FUNCTIONAL TEST 

contained in TS 3.3.1.1 from 7 days to 

31 days. 

Date of issuance: April 27, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment No.: 141. 

Facility Operating License No. NPF–

58: This amendment revised the 

Technical Specifications and License. 

Date of initial notice in Federal 

Register: March 28, 2006 (71 FR 15484) 

The October 17, 2006 and February 8, 

2007 supplements, 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 amendment is contained in a 

Safety Evaluation dated April 27, 2007. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit No. 1, 

Lake County, Ohio 

Date of application for amendment: 

November 21, 2005, as supplemented by 

letter dated February 22, 2007. 

Brief description of amendment: This 

amendment revised the acceptance 

criteria of technical specification (TS) 

surveillance requirements associated 

with TS 3.8.1, to modify the emergency 

diesel generator start tests to provide 

minimum voltage and frequency limits 

and clarified other limits as steady state 

parameters. 


jlentini on PROD1PC65 with NOTICES
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Date of issuance: April 30, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days. 

Amendment No.: 142. 

Facility Operating License No. NPF–

58: This amendment revised the 

Technical Specifications and License. 

Date of initial notice in Federal 

Register: January 17, 2006 (71 FR 2591) 

The February 22, 2007, 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 amendment is contained in a 

Safety Evaluation dated April 30, 2007. 

No significant hazards consideration 

comments received: No. 

Florida Power and Light Company, 

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

Point Plant, Units 3 and 4, Miami-Dade 

County, Florida 

Date of application for amendments: 

April 27, 2006, as supplemented 

December 5, 2006 and March 1, 2007. 

Brief description of amendments: 

These amendments revised the existing 

steam generator tube surveillance 

program to be consistent with the 

Technical Specification Task Force 

(TSTF) Standard TS Change Traveler, 

TSTF–449, ‘‘Steam Generator Tube 

Integrity.’’

Date of issuance: April 27, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment Nos: 233 and 228. 

Renewed Facility Operating License 

Nos. DPR–31 and DPR–41: Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: July 18, 2006 (71 FR 40748). 

The supplements dated December 5, 

2006, and March 1, 2007, provided 

additional information 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 April 27, 2007. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket Nos. 50–266 and 50–301, Point 

Beach Nuclear Plant, Units 1 and 2, 

Town of Two Creeks, Manitowoc 

County, Wisconsin 

Date of application for amendments: 

June 6, 2006. 

Brief description of amendments: The 

amendments revise information in the 

Final Safety Analysis Report (FSAR) 

regarding the reactor pressure vessel 

Charpy upper shelf energy (USE) 

requirements of Title 10 of the Code of 

Federal Regulations Part 50, Appendix 

G, Section IV.A.1.c. The change updates 

the analysis for satisfying the RPV 

Charpy USE requirements through the 

end of the current operating licenses. 

Date of issuance: May 10, 2007. 

Effective date: As of the date of 

issuance and shall be incorporated into 

the FSAR during the next update of the 

FSAR, as required by 10 CFR 50.71(c). 

Amendment Nos.: 227 and 232. 

Renewed Facility Operating License 

Nos. DPR–24 and DPR–27: Amendments 

revise the Final Safety Analysis Report 

and the Licenses. 

Date of initial notice in Federal 

Register: July 18, 2006 (71 FR 40750). 

The Commission’s related evaluation 

of the amendments is contained in a 

safety evaluation dated May 10, 2007. 

No significant hazards consideration 

comments received: No. 

PPL Susquehanna, LLC, Docket No. 50–

387 and 50–388, Susquehanna Steam 

Electric Station, Units 1 and 2 (SSES 1 

and 2), Luzerne County, Pennsylvania 

Date of application for amendments: 

May 31, 2006. 

Brief description of amendments: The 

amendments correct administrative 

errors in the SSES 1 and 2 Technical 

Specifications (TSs) by adding a logical 

‘‘AND’’ connector in Condition B of TS 

3.8.1 for SSES 1, ‘‘AC Sources—

Operating,’’ and correct the routing of 

Interstate Route 80 on Figure 4.1–2 of 

TSs 4.1.2, ‘‘Low Population Zone,’’ for 

SSES 1 and 2. 

Date of issuance: April 26, 2007. 

Effective date: As of the date of 

issuance and to be implemented within 

30 days. 

Amendment Nos.: 243 and 221. 

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments 

revised the TSs and License. 

Date of initial notice in Federal 

Register: December 19, 2006 (71 FR 

75996). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated April 26, 2007. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket 

Nos. 50–259, 50–260, and 50–296, 

Browns Ferry Nuclear Plant, Units 1, 2, 

and 3, Limestone County, Alabama 

Date of application for amendments: 

November 15, 2006 January 11, 2007, as 

supplemented by letters dated January 

11, and April 24, 2007. 

Description of amendment request: 

The amendments revised the Fire 

Protection License Condition numbers 

(13), (14), and (7) for Units 1, 2, and 3, 

respectively, to accommodate operation. 

Date of issuance: April 25, 2007. 

Effective date: Date of issuance, to be 

implemented within 30 days. 

Amendment Nos.: 271, 300, and 259. 

Renewed Facility Operating License 

Nos. DPR–33, DPR–52, and DPR–68: 

Amendments revised the Operating 

Licenses. 

Date of initial notice in Federal 

Register: December 19, 2006 (71 FR 

76000). The supplements dated January 

11, and April 24, 2007, 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 amendments is contained in a 

Safety Evaluation dated April 25, 2007. 

No significant hazards consideration 

comments received: No. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Unit Nos. 

1 and 2, Somervell County, Texas 

Date of amendment request: March 

22, 2006, supplemented by letter dated 

September 12, 2006. 

Brief description of amendments: The 

amendments revised the Technical 

Specification (TS) 3.8.1 entitled, ‘‘AC 

Sources—Operating.’’ Specifically, the 

proposed change would revise the 

completion time for TS 3.8.1, Condition 

F, Required Action F.1 from 12 hours to 

24 hours. 

Currently, TS 3.8.1, Condition F 

requires that an inoperable safety 

injection (SI) sequencer must be 

restored to operable status within 12 

hours. If this completion time is not 

met, Condition G becomes applicable 

and the plant must be shutdown to at 

least Mode 3 within the following 6 

hours. The proposed change to the 

completion time for TS 3.8.1, Condition 

F, Required Action F.1 provides more 

time to complete necessary repairs and 

required post-work testing to restore an 

inoperable SI sequencer to operable 

status prior to commencing a plant 

shutdown to Mode 3. 

Date of issuance: April 27, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: NPF–87—138, 

NPF–89—138. 

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments 


jlentini on PROD1PC65 with NOTICES
28728 Federal Register / Vol. 72, No. 98 / Tuesday, May 22, 2007 / Notices 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: March 28, 2007 (72 FR 

14623). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated April 27, 2007. 

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: 

June 7, 2006. 

Brief description of amendment: The 

amendment deleted Required Action 

D.1.2 in Technical Specification (TS) 

3.7.10, ‘‘Control Room Emergency 

Ventilation System (CREVS),’’ and 

Required Action C.1.2 in TS 3.7.11, 

‘‘Control Room Air Conditioning System 

(CRACS).’’ For TS 3.7.13, ‘‘Emergency 

Exhaust System (EES),’’ the amendment 

also deletes the phrase ‘‘in MODE 1, 2, 

3, or 4’’ from Condition A (one EES train 

inoperable) and revised Condition D to 

state the following: ‘‘Required Action 

and associated Completion Time of 

Condition A not met during movement 

of irradiated fuel assemblies in the fuel 

building.’’

Date of issuance: May 9, 2007. 

Effective date: As of its date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 184. 

Facility Operating License No. NPF–

30: The amendment revised the 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 1, 2006 (71 FR 43536) 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated May 9, 2007. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company, et 

al., Docket Nos. 50–280 and 50–281, 

Surry Power Station, Units 1 and 2, 

Surry County, Virginia 

Date of application for amendments: 

January 31, 2007. 

Brief Description of amendments: 

These amendments revised the 

Technical Specification surveillance 

requirements for addressing a missed 

surveillance, and is consistent with the 

Nuclear Regulatory Commission 

approved Revision 6 of Technical 

Specification Task Force (TSTF) 

Standard Technical Specifications 

Change Traveler TSTF–358, ‘‘Missed 

Surveillance Requirements.’’

Date of issuance: May 3, 2007. 

Effective date: As of date of issuance 

and shall be implemented within 30 

days. 

Amendment Nos.: 253, 252. 

Renewed Facility Operating License 

Nos. DPR–32 and DPR–37: Amendments 

changed the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: February 27, 2007 (72 FR 

8806). 

No significant hazards consideration 

comments received: No. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated May 3, 2007. 

Dated at Rockville, Maryland, this 11th day 

of May, 2007. 

For the Nuclear Regulatory Commission. 

Timothy McGinty, 

Acting Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E7–9523 Filed 5–21–07; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Draft Supplements to Revision 9 of 

NUREG–1021, ‘‘Operator 

LicensingExamination Standards for 

Power Reactors,’’ and to Revision 2 of 

NUREG–1122 [and –1123] ‘‘Knowledge 

and Abilities Catalog for Nuclear 

Power Plant Operators: Pressurized 

[Boiling] Water Reactors’’

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Notice of proposed supplements 

for public comment. 

SUMMARY: The Nuclear Regulatory 

Commission (NRC) has issued for public 

comment draft supplements to Revision 

9 of NUREG–1021, ‘‘Operator Licensing 

Examination Standards for Power 

Reactors,’’ and to Revision 2 of NUREG–

1122 [and –1123] ‘‘Knowledge and 

Abilities Catalog for Nuclear Power 

Plant Operators: Pressurized [Boiling] 

Water Reactors.’’ These NUREGs 

provide policy and guidance for the 

development, administration, and 

grading of examinations used for 

licensing operators at nuclear power 

plants pursuant to the Commission’s 

regulations in 10 CFR Part 55, 

‘‘Operators’’ Licenses.’’ NUREG–1021 

also provides guidance for maintaining 

operators’ licenses, and for the NRC to 

conduct requalification examinations, 

when necessary. 

The draft supplement to Revision 9 of 

NUREG–1021 includes a number of 

minor changes that are intended to: (1) 

Clarify licensed operator medical 

requirements, including the use of 

prescription medications; (2) clarify the 

use of surrogate operators during 

dynamic simulator scenarios; (3) clarify 

the selection process for generic 

knowledge and ability (K/A) statements; 

(4) qualify the NRC review of postexamination comments; (5) provide 

additional guidance for maintaining an 

active license (watchstander 

proficiency) and license reactivation; 

and (6) conform with proposed updates 

to NUREGs–1122 and –1123, which are 

concurrently available for public 

comment. The proposed changes are 

summarized in the Record of Proposed 

Changes, and identified by highlight/ 

redline and strikeouts. 

The draft supplements to NUREGs–

1122 and –1123 propose to reword and 

reorganize Section 2, ‘‘Generic 

Knowledge and Abilities,’’ and add a 

new K/A topic to Section 4, 

‘‘Emergency/Abnormal Plant 

Evolutions,’’ to address generator 

voltage and electric grid disturbances. 

The proposed changes are summarized 

in the Record of Changes, and identified 

by highlight/redline and strikeouts. 

Availability: The draft supplements 

are available electronically via the 

NRC’s Public Electronic Reading Room 

(http://www.nrc.gov/public-involve/doccomment.html) and in the NRC’s Public 

Document Room located at 11555 

Rockville Pike, Rockville, Maryland. If 

you do not have electronic access to 

NRC documents, single copies of the 

draft supplements are available upon 

request, by contacting David S. Muller 

by phone at (301) 415–1412 or by e-mail 

at dsm3@nrc.gov. 

DATES: Comments must be provided by 

July 23, 2007. Comments received after 

this date will be considered if 

practicable to do so, but only those 

comments received on or before the due 

date can be assured consideration. 

ADDRESSES: Submit written comments 

to the Chief, Rules, Directives, and 

Editing Branch, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Mail Stop T6–D59, 

Washington, DC 20555–0001, and 

specify the report number in your 

comments. You may also provide 

comments via the NRC’s Public 

Electronic Reading Room by following 

the instructions at http://www.nrc.gov/ 

public-involve/doc-comment/form.html. 

FOR FURTHER INFORMATION CONTACT: 

David S. Muller, Operator Licensing and 

Human Performance Branch, Office of 

Nuclear Reactor Regulation, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–001. Telephone: 

(301) 415–1412; e-mail: dsm3@nrc.gov. 


jlentini on PROD1PC65 with NOTICES