Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 14303 

Week of April 30, 2007—Tentative 

There are no meetings scheduled for 

the Week of April 30, 2007. 

* * * * * 

*The schedule for Commission 

meetings is subject to change on short 

notice. To verify the status of meetings 

call (recording)—(301) 415–1292. 

Contact person for more information: 

Michelle Schroll, (301) 415–1662. 

* * * * * 

Additional Information 

Affirmation of ‘‘Consumers Energy 

Company, et al. (Palisades Nuclear 

Plant); License Transfer Application’’

tentatively scheduled on Thursday, 

March 22, 2007, has been tentatively 

rescheduled on Thursday, March 29, 

2007, at 9:25 a.m. 

* * * * * 

The NRC Commission Meeting 

Schedule can be found on the Internet 

at: www.nrc.gov/about-nrc/policymaking/schedule.html. 

* * * * * 

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g., 

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

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: March 22, 2007. 

R. Michelle Schroll, 

Office of the Secretary. 

[FR Doc. 07–1501 Filed 3–23–07; 12:25 pm] 

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 March 2, 

2007 to March 15, 2007. The last 

biweekly notice was published on 

March 13, 2007 (72 FR 11383). 

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 


sroberts on PROD1PC70 with NOTICES
14304 Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

A request for a hearing or a petition 

for leave to intervene must be filed by: 

(1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, Attention: Rulemaking and 

Adjudications Staff; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland, 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) E-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HearingDocket@nrc.gov; or (4) facsimile 

transmission addressed to the Office of 

the Secretary, U.S. Nuclear Regulatory 

Commission, Washington, DC, 

Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

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

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer of 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(i)–(viii). 

For further details with respect to this 

action, see the application for 

amendment which is available for 

public inspection at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the ADAMS Public 

Electronic Reading Room on the Internet 

at the NRC Web site, http:// 

www.nrc.gov/reading-rm/adams.html. If 

you do not have access to ADAMS or if 

there are problems in accessing the 

documents located in ADAMS, contact 

the PDR Reference staff at 1 (800) 397–

4209, (301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

Entergy Operations, Inc., System Energy 

Resources, Inc., South Mississippi 

Electric Power Association, and Entergy 

Mississippi, Inc. 

Docket No. 50–416, Grand Gulf 

Nuclear Station, Unit 1, Claiborne 

County, Mississippi. 

Date of amendment request: February 

8, 2007. 

Description of amendment request: 

The proposed amendment would 

modify Grand Gulf Nuclear Station, 

Unit 1 (GGNS) technical specification 

(TS) requirements for MODE change 

limitations in limiting condition for 

operation (LCO) 3.0.4 and surveillance 

requirement (SR) 3.0.4. The proposed 

TS changes are consistent with Revision 

9 of Nuclear Regulatory Commission 

(NRC) approved Industry TS Task Force 

(TSTF) Standard TS Change Traveler, 

TSTF–359, ‘‘Increase Flexibility in 

MODE Restraints.’’ In addition, the 

proposed amendment would also 

change TS Section 1.4, Frequency, 

Example 1.4–1, ‘‘Surveillance 

Requirements,’’ to accurately reflect the 

changes made by TSTF–359, which is 

consistent with NRC-approved TSTF–

485, Revision 0, ‘‘Correct Example 1.4–

1.’’

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on August 2, 2002 (67 FR 

50475), as part of the Consolidated Line 

Item Improvement Process (CLIIP), on 

possible amendments to revise the 


sroberts on PROD1PC70 with NOTICES
Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 14305 

plant-specific TS to modify 

requirements for MODE change 

limitations in LCO 3.0.4 and SR 3.0.4. 

The NRC staff subsequently issued a 

notice of availability of the models for 

Safety Evaluation and No Significant 

Hazards Consideration Determination 

for referencing in license amendment 

applications in the Federal Register on 

April 4, 2003 (68 FR 16579). The 

licensee affirmed the applicability of the 

CLIIP, including the model No 

Significant Hazards Consideration 

Determination, in its application dated 

February 8, 2007. 

The proposed TS changes are 

consistent with NRC-approved Industry 

TSTF Standard TS change, TSTF–359, 

Revision 8, as modified by 68 FR 16579. 

TSTF–359, Revision 8, was 

subsequently revised to incorporate the 

modifications discussed in the April 4, 

2003, Federal Register notice and other 

minor changes. TSTF–359, Revision 9, 

was subsequently submitted to the NRC 

on April 28, 2003, and was approved by 

the NRC on May 9, 2003. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

NRC staff’s analysis of the issue of no 

significant hazards consideration is 

presented below: 

Criterion 1—The Proposed Changes Do 

Not Involve a Significant Increase in the 

Probability or Consequences of an 

Accident Previously Evaluated 

The proposed changes in TS Section 

1.4, Frequency, Example 1.4–1, would 

accurately reflect the changes made by 

TSTF–359 in LCO 3.0.4 and SR 3.0.4, 

which are consistent with NRCapproved TSTF–485, Revision 0. These 

changes are considered administrative 

in that they modify the example to 

demonstrate the proper application of 

LCO 3.0.4 and SR 3.0.4. The 

requirements of LCO 3.0.4 and SR 3.0.4 

are clear and are clearly explained in 

the associated Bases. As a result, 

modifying the example will not result in 

a change in usage of the TS. 

The proposed changes in LCO 3.0.4 

and SR 3.0.4 allow entry into a mode or 

other specified condition in the 

applicability of a TS, while in a TS 

condition statement and the associated 

required actions of the TS. The 

proposed changes do not adversely 

affect accident initiators or precursors, 

the ability of structures, systems, and 

components to perform their intended 

function to mitigate the consequences of 

an initiating event within the assumed 

acceptance limits, or radiological release 

assumptions used in evaluating the 

radiological consequences of an 

accident previously evaluated. Being in 

a TS condition and the associated 

required actions are not an initiator of 

any accident previously evaluated. 

Therefore, the probability of an accident 

previously evaluated is not significantly 

increased. The consequences of an 

accident while relying on required 

actions as allowed by proposed LCO 

3.0.4, are no different than the 

consequences of an accident while 

entering and relying on the required 

actions while starting in a condition of 

applicability of the TS. Therefore, the 

consequences of an accident previously 

evaluated are not significantly affected 

by these changes. The addition of a 

requirement to assess and manage the 

risk introduced by these changes will 

further minimize possible concerns. 

Therefore, these changes do not involve 

a significant increase in the probability 

or consequences of an accident 

previously evaluated. 

Criterion 2—The Proposed Changes Do 

Not Create the Possibility of a New or 

Different Kind of Accident from any 

Previously Evaluated 

No new or different accidents result 

from utilizing the proposed changes. 

The proposed changes do not involve a 

physical alteration of the plant (no new 

or different type of equipment will be 

installed) or a change in the methods 

governing normal plant operation. In 

addition, the changes do not impose any 

new or different requirements or 

eliminate any existing requirements. 

The proposed changes do not alter 

assumptions made in the safety analysis 

and are consistent with the safety 

analysis assumptions and current plant 

operating practice. Entering into a mode 

or other specified condition in the 

applicability of a TS, while in a TS 

condition statement and the associated 

required actions of the TS, will not 

introduce new failure modes or effects 

and will not, in the absence of other 

unrelated failures, lead to an accident 

whose consequences exceed the 

consequences of accidents previously 

evaluated. The addition of a 

requirement to assess and manage the 

risk introduced by these changes will 

further minimize possible concerns. 

Thus, these changes do not create the 

possibility of a new or different kind of 

accident from an accident previously 

evaluated. 

Criterion 3—The Proposed Changes Do 

Not Involve a Significant Reduction in 

the Margin of Safety 

The proposed changes in TS Section 

1.4, Example 1.4–1, are considered 

administrative and will have no effect 

on the application of the TS 

requirements. Therefore, the margin of 

safety provided by the TS requirements 

is unchanged. The proposed changes in 

TS LCO 3.0.4 and SR 3.0.4 allow entry 

into a mode or other specified condition 

in the applicability of a TS, while in a 

TS condition statement and the 

associated required actions of the TS. 

The GGNS TS allows operation of the 

plant without the full complement of 

equipment through the TS conditions 

for not meeting the TS LCO. The risk 

associated with this allowance is 

managed by the imposition of required 

actions that must be performed within 

the prescribed completion times. The 

net effect of being in a TS LCO 

condition on the margin of safety is not 

considered significant. The proposed 

changes do not alter the required actions 

or completion times of the TS. The 

proposed changes allow TS conditions 

to be entered, and the associated 

required actions and completion times 

to be used in new circumstances. This 

use is predicated upon the licensee’s 

performance of a risk assessment and 

the management of plant risk. The 

changes also eliminate current 

allowances for utilizing required actions 

and completion times in similar 

circumstances, without assessing and 

managing risk. The net change to the 

margin of safety is insignificant. 

Therefore, these changes do not involve 

a significant reduction in a margin of 

safety. 

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Terence A. 

Burke, Associate General Council—

Nuclear Entergy Services, Inc., 1340 

Echelon Parkway, Jackson, Mississippi 

39213. 

NRC Branch Chief: David Terao. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2, 

Ogle County, Illinois 

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois. 

Date of amendment request: January 

8, 2007. 

Description of amendment request: 

The proposed amendment would revise 

the technical specification (TS) 

requirements for selected reactor trip 

system (RTS) instrumentation, 

engineered safety feature actuation 

system (ESFAS) instrumentation, and 

containment ventilation isolation 

instrumentation to adopt completion 

times, test bypass time, and surveillance 

test interval changes. The changes are 

based on Westinghouse Electric 

Company, LLC, topical reports WCAP–

14333–P–A, Revision 1, ‘‘Probabilistic 


sroberts on PROD1PC70 with NOTICES
14306 Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 

Risk Analysis of the [Reactor Protection 

System] RPS and ESFAS Test Times and 

Completion Times,’’ and WCAP–15376–

P–A, Revision 1, ‘‘Risk-Informed 

Assessment of the RTS and ESFAS 

Surveillance Test Intervals and Reactor 

Trip Breaker Test and Completion 

Times.’’

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Overall protection system performance will 

remain within the bounds of the previously 

performed accident analyses since no 

hardware changes are proposed. The same 

RTS and ESFAS instrumentation will 

continue to be used. The protection systems 

will continue to function in a manner 

consistent with the plant design basis. These 

changes to the TS do not result in a condition 

where the design, material, and construction 

standards that were applicable prior to the 

change are altered. 

The proposed changes will not modify any 

system interface. The proposed changes will 

not affect the probability of any event 

initiators. There will be no degradation in the 

performance of or an increase in the number 

of challenges imposed on safety-related 

equipment assumed to function during an 

accident situation. There will be no change 

to normal plant operating parameters or 

accident mitigation performance. The 

proposed changes will not alter any 

assumptions or change any mitigation actions 

in the radiological consequence evaluations 

in the Updated Final Safety Analysis Report. 

The determination that the results of the 

proposed changes are acceptable was 

established in the NRC Safety Evaluations 

prepared for WCAP–14333–P–A, (issued by 

letter dated July 15, 1998) and for WCAP–

15376–P–A, (issued by letter dated December 

20, 2002). Implementation of the proposed 

changes will result in an insignificant risk 

impact. 

Applicability of these conclusions has been 

verified through plant-specific reviews and 

implementation of the generic analysis 

results in accordance with the respective 

NRC Safety Evaluation conditions. 

The proposed changes to the CTs 

[completion times], test bypass times, and 

Surveillance Frequencies reduce the 

potential for inadvertent reactor trips and 

spurious engineered safeguard features 

actuations, and therefore do not increase the 

probability of any accident previously 

evaluated. The proposed changes do not 

change the response of the plant to any 

accidents and have an insignificant impact 

on the reliability of the RTS and ESFAS 

signals. The RTS and ESFAS will remain 

highly reliable and the proposed changes will 

not result in a significant increase in the risk 

of plant operation. This is demonstrated by 

showing that the impact on plant safety, as 

measured by the increase in core damage 

frequency (CDF) is less than 1.0E–06 per year 

and the increase in large early release 

frequency (LERF) is less than 1.0E–07 per 

year. In addition, for the CT changes, the 

incremental conditional core damage 

probabilities (ICCDP) and incremental 

conditional large early release probabilities 

(ICLERP) are less than 5.0E–07 and 5.0E–08, 

respectively. These changes meet the 

acceptance criteria in Regulatory Guides 

(RGs) 1.174 and 1.177. Therefore, since the 

RTS and ESFAS will continue to perform 

their functions with high reliability, as 

originally assumed, and the increase in risk, 

as measured by DCDF, DLERF, ICCDP, 

ICLERP risk metrics, is within the acceptance 

criteria of existing regulatory guidance, there 

will not be a significant increase in the 

consequences of any accidents. 

The proposed changes do not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, or 

configuration of the facility or the manner in 

which the plant is operated and maintained. 

The proposed changes do not alter or prevent 

the ability of structures, systems, and 

components from performing their intended 

function to mitigate the consequences of an 

initiating event within the assumed 

acceptance limits. The proposed changes do 

not affect the source term, containment 

isolation, or radiological release assumptions 

used in evaluating the radiological 

consequences of any accident previously 

evaluated. The proposed changes are 

consistent with safety analysis assumptions 

and resultant consequences. 

Therefore, this change does not increase 

the probability or consequences of any 

accident previously evaluated. 

2. The proposed changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. There are no hardware changes 

nor are there any changes in the method by 

which any safety-related plant system 

performs its safety function. The proposed 

changes will not affect the normal method of 

plant operation. No performance 

requirements will be affected or eliminated. 

The proposed changes will not result in 

physical alteration to any plant system nor 

will there be any change in the method by 

which any safety-related plant system 

performs its safety function. There will be no 

setpoint changes or changes to accident 

analysis assumptions. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures are introduced as a result of 

these changes. There will be no adverse effect 

or challenges imposed on any safety-related 

system as a result of these changes. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. The proposed changes do not involve a 

significant reduction in a margin of safety? 

The proposed changes do not affect the 

acceptance criteria for any analyzed event 

nor is there a change to any Safety Analysis 

Limit. There will be no effect on the manner 

in which safety limits, limiting safety system 

settings, or limiting conditions for operation 

are determined nor will there be any effect 

on those plant systems necessary to assure 

the accomplishment of protection functions. 

There will be no impact on the departure 

from nucleate boiling limits, fuel centerline 

temperature, or any other margin of safety. 

The radiological dose consequence 

acceptance criteria listed in the NUREG–

0800, ‘‘Standard Review Plan for the Review 

of Safety Analysis Reports for Nuclear Power 

Plants,’’ will continue to be met. 

Redundant RTS and ESFAS trains are 

maintained, and diversity with regard of the 

signals that provide reactor trip and 

engineered safety features actuation is also 

maintained. All signals credited as primary 

or secondary, and all operator actions 

credited in the accident analyses will remain 

the same. The proposed changes will not 

result in plant operation in a configuration 

outside the design basis. The calculated 

impact on risk is insignificant and meets the 

acceptance criteria contained in RGs 1.174 

and 1.177. Although there was no attempt to 

quantify any positive human factors benefit 

due to increased CTs and bypass test times, 

it is expected that there would be a net 

benefit due to a reduced potential for 

spurious reactor trips and actuations 

associated with testing. 

Implementation of the proposed changes is 

expected to result in an overall improvement 

in safety, as follows: 

• Reduced testing will result in fewer 

inadvertent reactor trips, less frequent 

actuation of ESFAS components, less 

frequent distraction of operations personnel 

without significantly affecting RTS and 

ESFAS reliability. 

• Improvements in the effectiveness of the 

operating staff in monitoring and controlling 

plant operation will be realized. This is due 

to less frequent distraction of the operators 

and shift supervisor to attend to 

instrumentation Required Actions with short 

CTs. 

• Longer repair times associated with 

increased CTs will lead to higher quality 

repairs and improved reliability. 

• The CT extensions for the reactor trip 

breakers will provide additional time to 

complete test and maintenance activities 

while at power, potentially reducing the 

number of forced outages related to 

compliance with reactor trip breaker CT, and 

provide consistency with the CT for the logic 

trains. 

Therefore, the proposed changes do 

not involve a significant reduction in 

the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Russell A. Gibbs. 


sroberts on PROD1PC70 with NOTICES
Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 14307 

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant, Wright County, 

Minnesota 

Date of amendment request: January 

30, 2007. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specifications (TSs) 

Surveillance Requirement (SR) 3.5.1.3.b 

to correctly state that the required 

pressure at which the Alternate 

Nitrogen System is determined to be 

operable should be greater than or equal 

to 410 psig, not the currently stated 

pressure of greater than or equal to 220 

psig. The safety-related Alternate 

Nitrogen System provides an alternate 

pressure source to equipment required 

during or following an accident. The 

licensee has determined that the current 

acceptance value specified by SR 

3.5.1.3.b is non-conservative and needs 

to be corrected to the higher value. 

Basis for proposed no significant 

hazards consideration determination: 

As required by Title 10 of the Code of 

Federal Regulations (10 CFR) Part 

50.91(a), the licensee has provided its 

analysis of the issue of no significant 

hazards consideration (NSHC). The NRC 

staff reviewed the licensee’s analysis, 

and has performed its own analysis as 

follows: 

(1) Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed amendment would only 

correct the acceptance value specified by SR 

3.5.1.3.b. The acceptance value of the 

nitrogen supply was not considered to be a 

precursor to, and does not affect the 

probability of, an accident. In addition, there 

is no design or operation change associated 

with the proposed amendment. 

Therefore, the proposed amendment does 

not increase the probability of an accident 

previously evaluated. 

The corrected, higher pressure of the 

Alternate Nitrogen System will ensure that 

nitrogen is available to operate equipment 

after an accident, as designed. The increased 

acceptance value will not decrease the 

functionality of the Alternate Nitrogen 

System, or the functionality of the plant 

equipment it supports. Therefore, the plant 

systems required to mitigate accidents will 

remain capable of performing their design 

functions. As a result, the proposed 

amendment will not lead to a significant 

change in the consequences of any accident. 

(2) Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. The proposed amendment does not 

involve a physical alteration of any system, 

structure, or component (SSC) or a change in 

the way any SSC is operated. The proposed 

amendment does not involve operation of 

any SSCs in a manner or configuration 

different from those previously recognized or 

evaluated. No new failure mechanisms will 

be introduced by the revised acceptance 

value. 

Thus, the proposed amendment does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

(3) Does the proposed amendment involve 

a significant reduction in a margin of safety? 

No. The proposed amendment only 

changes the acceptance value of the Alternate 

Nitrogen System. There will be no 

modification of any TSs limiting condition 

for operation, no change to any limit on 

previously analyzed accidents, no change to 

how previously analyzed accidents or 

transients would be mitigated, no change in 

any methodology used to evaluate 

consequences of accidents, and no change in 

any operating procedure or process. 

Therefore, the proposed amendment does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on the 

NRC staff’s own analysis above, it 

appears that the three standards of 10 

CFR 50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

proposed amendment involves no 

significant hazards consideration. 

Attorney for licensee: Jonathan Rogoff, 

Esquire, Vice President, Counsel & 

Secretary, Nuclear Management 

Company, LLC, 700 First Street, 

Hudson, WI 54016. 

NRC Branch Chief: L. Raghavan. 

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: 

February 8, 2007. 

Description of amendment requests: 

This license amendment request will (1) 

revise Technical Specification (TS) 

Surveillance Requirement (SR) 3.3.7.3.a 

to lower the allowable value for dropout 

and raise the allowable value for pickup 

of the degraded voltage function, and (2) 

revise TS SR 3.8.1 to lower the diesel 

generator minimum output voltage due 

to lower settings for the degraded 

voltage function. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This proposed change revises the 

Technical Specification (TS) Surveillance 

Requirement (SR) 3.3.7.3.a allowable values 

of the Degraded Voltage Function and SRs 

3.8.1.2, .7, .9, .11, .12, .15, .16, .17, .19, and 

.20 for Diesel Generator (DG) minimum 

operable voltage. This proposed change will 

allow Southern California Edison (SCE) to 

widen the operating band while maintaining 

adequate conservatism for the degraded relay 

settings and overall loop uncertainties while 

keeping 218 kV as the minimum voltage on 

the offsite transmission grid necessary to 

support operability of the immediate access 

offsite power source (also referred to as the 

normal preferred power source). This will be 

accomplished by lowering the dropout and 

increasing the pickup settings of the 

degraded voltage protection relays. Following 

approval of this proposed change, the 4.16 

kV Class 1E buses would remain on the 

normal preferred power source at or above a 

grid voltage of 218 kV while protecting all 

Class 1E equipment from degraded grid 

conditions. 

The degraded voltage protection circuits 

are designed to protect electrical equipment 

against the effects of degraded voltage on the 

offsite transmission networks. Therefore, 

these circuits are generally not considered to 

be accident initiators. However, spurious 

actuation of the degraded voltage protection 

relays could result in the loss of the preferred 

power source (offsite source of alternating 

current (AC) power). The proposed change 

lowers the allowable value for dropout and 

raises the allowable value for pickup for the 

degraded voltage protection relays. This 

results in an increase in operating band and 

a lower probability of spurious actuation of 

these degraded voltage signals. Therefore, 

there is no increase in the probability of a 

Loss of Offsite Power (preferred power 

source) as a result of this proposed change. 

The safety function of the degraded voltage 

protection circuits is to ensure the operability 

of Class 1E equipment. SCE has performed 

calculations that demonstrate that operation 

in accordance with this proposed change will 

not result in operation of plant equipment at 

degraded voltages. Therefore, there is no 

increase in the consequences of any accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed allowable values of the 

degraded voltage relays and the DG 

minimum operating voltage will provide an 

acceptable level of protection for plant 

equipment. 

This proposed change affects only the 

voltage settings of the degraded voltage 

protection relays and voltage regulator setting 

of the DG for lowering the required bus 

voltage. There is no other change to the 

degraded voltage function. There are no 

physical modifications necessary to the 

degraded voltage protection relays or the DG. 

There are no changes to the actions 

performed by the relays or the DG following 

actuation. Therefore, there are no new failure 


sroberts on PROD1PC70 with NOTICES
14308 Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 

modes or effects introduced by this proposed 

change. 

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 degraded voltage protection 

schemes are designed to ensure that plant 

equipment will not operate at a degraded 

voltage and the DG Automatic Voltage 

Regulator (AVR) is set to provide adequate 

voltage for resetting of the relays and 

satisfactory operation of the Safety Related 

equipment. The proposed degraded voltage 

allowable values will not affect the existing 

protection criterion for plant equipment. This 

maintains the existing margin of safety for 

plant equipment. 

Therefore, there is no significant reduction 

in a margin of safety as a result of the 

proposed amendment. 

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: David Terao. 

Southern Nuclear Operating Company, 

Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of Georgia, 

City of Dalton, Georgia 

Docket Nos. 50–321 and 50–366, 

Edwin I. Hatch Nuclear Plant, Units 1 

and 2, Appling County, Georgia. 

Date of amendment request: February 

13, 2007. 

Description of amendment request: 

The proposed amendment would 

modify the licensee’s Technical 

Specification (TS) Section 3.9.1, 

‘‘Refueling Equipment Interlocks,’’ to 

add required actions to allow insertion 

of a control rod withdrawal block and 

verification that all control rods are 

fully inserted as alternate actions to 

suspending in-vessel fuel movement in 

the event that one or more required 

refueling equipment interlocks are 

inoperable. These changes are based on 

Technical Specification Task Force 

(TSTF) change TSTF–225, Revision 2, 

‘‘Fuel movement with inoperable 

refueling equipment interlocks’’ and are 

consistent with the current Boiling 

Water Reactor (BWR)/4 Standard 

Technical Specifications (STS), 

NUREG–1433, Volume 1, Revision 3.0. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

The proposed change provides additional 

actions for an inoperable refueling equipment 

interlock. The proposed actions will allow 

fuel movement with inoperable refueling 

interlocks, however, those actions will 

require the insertion of a continuous control 

rod withdrawal block, as well as verification 

that all control rods are fully inserted, before 

the commencement of fuel movement. Since 

fuel movement with the refueling interlocks 

operable allows control rod withdrawal 

under some circumstances, complete 

prevention of control rod withdrawal with 

the refueling interlocks inoperable does not 

increase the likelihood of a reactivity event, 

and may in fact decrease its probability of 

occurrence. 

The refueling interlocks are not designed 

or otherwise intended to prevent or mitigate 

the consequences of the fuel handling 

accident. This proposed change does not 

involve those structures that could have an 

effect on the fuel handling accident and its 

consequences, such as the fuel design, the 

integrity of the refueling platform, and the 

integrity of the refueling mast and grapple. 

Furthermore, the consequences of the 

refueling accident are not increased since, 

should that accident occur while operating 

under the provisions of the alternate actions, 

all control rods will be fully inserted. The 

consequences of the fuel assembly insertion 

error event during refueling are not increased 

since this proposed change preserves the 

initial conditions of that transient event, i.e., 

all control rods inserted. 

Implementing these changes will not 

increase the likelihood of an equipment 

failure resulting from the use of the refueling 

cranes and hoists. Such protection is afforded 

by other plant (owner controlled) 

specifications and procedures. These 

documents require testing and maintenance 

of these components separate from the 

requirements of [Limiting Condition for 

Operation] LCO 3.9.1. 

This submittal does not affect any other 

system, structure or component that is 

important with respect to the prevention and 

mitigation of other accidents or transients. 

For the above reasons, this proposed 

Technical Specifications change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

The proposed change provides additional 

actions (the insertion of a control rod block 

and verification that all control rods are fully 

inserted) for inoperable refueling interlocks. 

This change does not involve any permanent 

alterations to plant systems or components. 

Nor does it involve changes to operational 

configurations or to the maintenance and 

testing of systems or components. 

Consequently, no new modes of operation are 

being introduced. Therefore, the 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 the margin of safety? 

The proposed change provides additional 

actions for an inoperable required refueling 

equipment interlock. The new actions will 

require that all control rods be fully inserted 

and that a control rod block be in effect. 

Under the current specifications, control rod 

withdrawal is allowed during fuel movement 

under certain conditions. 

The alternate actions of the proposed 

specifications will not allow rod withdrawal 

under any circumstances during fuel 

movement operations, therefore, this 

proposed change provides a level of safety at 

least equivalent to the existing actions. 

Consequently, the change does not involve 

a significant reduction in the margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ernest L. Blake, 

Jr., Esquire, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037. 

NRC Branch Chief: Evangelos C. 

Marinos. 

Virginia Electric and Power Company 

Docket Nos. 50–280 and 50–281, 

Surry Power Station, Unit Nos. 1 and 2, 

Surry County, Virginia. 

Date of amendment request: February 

26, 2007. 

Description of amendment request: 

The proposed change adds an operating 

license condition and revises the 

Technical Specifications to permit the 

replacement of main control room 

(MCR) and emergency switchgear room 

(ESGR) air-conditioning system (ACS) 

chilled water piping by using temporary 

45-day and 14-day allowed outage times 

(AOTs) four times in a 24-month time 

span. 

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 license amendment 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

The proposed change has been evaluated 

using the risk-informed processes described 


sroberts on PROD1PC70 with NOTICES
Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 14309 

in Regulatory Guide (RG) 1.174, ‘‘An 

Approach for Using Probabilistic Risk 

assessment in Risk-Informed Decisions on 

Plant-Specific Changes to the Licensing 

Basis,’’ and RG 1.177, ‘‘An Approach for 

Plant-Specific, Risk-Informed Decision 

Making: Technical Specifications.’’

The risk associated with the proposed 

change was found to be acceptably ‘‘small’’

and therefore not a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

In addition, the proposed change does not 

affect the initiators of analyzed events or the 

assumed mitigation of accident or transient 

events. During the temporary 45-day and 14- 

day AOT entries, equipment availability 

restrictions will restrict or limit the out-ofservice time of risk significant plant 

equipment due to surveillance testing, 

preventive maintenance, and elective 

maintenance. In addition, during the 

replacement activities, compensatory actions 

will be in place to ensure the availability of 

chilled water or to provide backup cooling. 

Therefore, the ACS will continue to perform 

its required function. As a result, the 

proposed change to the Surry TS does not 

involve any significant increase in the 

probability or the consequences of any 

accident or malfunction of equipment 

important to safety previously evaluated 

since neither accident probabilities nor 

consequences are being affected by this 

proposed change. 

2. Does the proposed license amendment 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated? 

The proposed change does not involve a 

change in the methods used to respond to 

plant transients. There is no alteration to the 

parameters within which the plant is 

normally operated or in the setpoints, which 

initiate protective or mitigative actions. The 

MCR and ESGR ACS will continue to 

perform its required function. This is assured 

by the planned implementation of 

compensatory actions, including provisions 

for backup cooling. Consequently, no new 

failure modes are introduced by the proposed 

change. Therefore, the proposed Surry TS 

change does not create the possibility of a 

new or different kind of accident or 

malfunction of equipment important to safety 

from any accident previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Margin of safety is established through the 

design of the plant structures, systems, and 

components, the parameters within which 

the plant is operated, and the establishment 

of the setpoints for the actuation of 

equipment relied upon to respond to an 

accident or transient event. The proposed 

change does not affect the ability of the MCR 

and ESGR ACS to perform its required 

function. This is assured by the planned 

implementation of compensatory actions, 

including provisions for backup cooling. 

Furthermore, the proposed change has been 

evaluated using the risk-informed processes 

described in Regulatory Guide (RG) 1.174, 

‘‘An approach for Using Probabilistic Risk 

Assessment in Risk-Informed Decisions on 

Plant-Specific Changes to the Licensing 

Basis,’’ and RG 1.177, ‘‘An Approach for 

Plant-Specific, Risk-Informed Decision 

Making: Technical Specifications.’’

The risk associated with the proposed 

change was found to be acceptably small. 

Therefore, the proposed change to the Surry 

TS 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: Lillian M. 

Cuoco, Esq., Senior Counsel, Dominion 

Resources Services, Inc., Millstone 

Power Station, Building 475, 5th Floor, 

Rope Ferry Road, Rt. 156, Waterford, 

Connecticut 06385. 

NRC Branch Chief: Evangelos C. 

Marinos. 

Virginia Electric and Power Company 

Docket Nos. 50–280 and 50–281, 

Surry Power Station, Unit Nos. 1 and 2, 

Surry County, Virginia. 

Date of amendment request: March 6, 

2007. 

Description of amendment request: 

The proposed amendments would 

revise the licensing basis (Updated Final 

Safety Analysis Report (UFSAR)) to 

permit irradiation of the fuel assemblies 

beginning with Surry Power Station, 

Unit Nos. 1 and 2, improved fuel 

assemblies with ZIRLO (Westinghouse 

trademark) cladding to a lead rod 

average burnup of 62,000 MWD/MTU. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The probability of occurrence or the 

consequence of an accident previously 

evaluated is not significantly increased. 

The activity being evaluated is a slight 

increase in the lead rod average burnup limit 

for the fuel assemblies. No change in fuel 

design or fuel enrichment will be required to 

increase the lead rod average burnup. The 

fuel rods at the extended lead rod average 

burnup will continue to meet the design 

limits with respect to fuel rod growth, clad 

fatigue, rod internal pressure and corrosion. 

There will be no impact on the capability to 

engage the fuel assemblies with the handling 

tools. Therefore, it is concluded that the 

change will not result in an increase in the 

probability of occurrence of any accident 

previously evaluated in the UFSAR. The 

impact of extending the lead rod average 

burnup to 62,000 MWD/MTU from 60,000 

MWD/MTU on the core kinetics parameter, 

core thermal-hydraulics/[departure from 

nucleate boiling ratio]DNBR, specific 

accident considerations, and radiological 

consequences was considered. Based on the 

evaluation of these considerations, it is 

concluded that increasing the lead rod 

average burnup limit to 62,000 MWD/MTU 

will not result in a significant increase in the 

consequences of the accidents previously 

evaluated in the Surry UFSAR. 

2. The possibility for a new or different 

type of accident from any accident 

previously evaluated is not created. 

The fuel is the only component affected by 

the change in the burnup limit. The change 

does not affect the thermal hydraulic 

response to any transient or accident. The 

existing fuel rod design criteria continue to 

be met at the higher burnup limit. Thus, the 

change does not create the possibility of an 

accident of a different type. 

3. The margin of safety as defined in the 

Bases to the Surry Technical Specifications is 

not significantly reduced. 

The operation of the Surry cores with a 

limited number of fuel assemblies with some 

fuel rods irradiated to a lead rod average 

burnup of 62,000 MWD/MTU will not change 

the performance requirements of any system 

or component such that any design criteria 

will be exceeded. The normal limits on core 

operation defined in the Surry Technical 

Specifications will remain applicable for the 

irradiation of the fuel to a lead rod average 

burnup of 62,000 MWD/MTU. Therefore, the 

margin of safety as defined in the Bases to 

the Surry Technical Specifications is not 

significantly reduced. 

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: Lillian M. 

Cuoco, Esq., Senior Counsel, Dominion 

Resources Services, Inc., Millstone 

Power Station, Building 475, 5th Floor, 

Rope Ferry Road, Rt. 156, Waterford, 

Connecticut 06385. 

NRC Branch Chief: Evangelos C. 

Marinos. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 


sroberts on PROD1PC70 with NOTICES
14310 Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

Florida Power Corporation, et al. 

Docket No. 50–302, Crystal River Unit 

3 Nuclear Generating Plant, Citrus 

County, Florida. 

Date of amendment request: February 

8, 2007. 

Description of amendment request: To 

change the basis for protection of spent 

fuel stored in the spent fuel pool (SFP) 

in order to eliminate the Final Safety 

Analysis Report commitment for 

maintaining the SFP missile shields. 

Date of publication of individual 

notice in the Federal Register: March 

13, 2007. (72 FR 11381). 

Expiration date of individual notice: 

May 14, 2007. 

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, 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

Carolina Power & Light Company 

Docket No. 50–261, H. B. Robinson 

Steam Electric Plant, Unit No. 2, 

Darlington County, South Carolina. 

Date of application for amendment: 

May 30, 2006, as supplemented by letter 

dated November 20, 2006. 

Brief description of amendment: The 

amendment revises the existing steam 

generator tube surveillance program at 

H. B. Robinson Steam Electric Plant, 

Unit No. 2. 

Date of issuance: March 12, 2007. 

Effective date: This license 

amendment is effective as of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No. 212. 

Renewed Facility Operating License 

No. DPR–23. Amendment revises the 

Technical Specifications. 

Date of initial notice in Federal 

Register: December 19, 2007 (71 FR 

75990). The November 20, 2006, 

supplemental letter provided clarifying 

information that did not change the 

initial proposed no significant hazards 

consideration determination. The 

Commission’s related evaluation of the 

amendment is contained in a Safety 

Evaluation dated: March 12, 2007. 

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 30, 2006, as supplemented by 

letter dated January 23, 2007. 

Brief description of amendment: The 

amendment modifies the radiological 

accident analyses and associated 

technical specifications. 

Date of issuance: March 8, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 190. 

Facility Operating License No. DPR–

43: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: March 14, 2006 (71 FR 

13172). 

The supplemental letter contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination, and did 

not expand the scope of the original 

Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 8, 2007. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc. 

Docket No. 50–423, Millstone Power 

Station, Unit No 3, New London County, 

Connecticut. 

Date of application for amendment: 

February 7, 2006, as supplemented by 

letters dated August 14, 2006, and 

January 2, 2007. 

Brief description of amendments: The 

amendment revised the Millstone Power 

Station, Unit No. 3 Technical 

Specifications to permit an increase in 

the allowed outage time from 72 hours 

to 7 days for the inoperablity of the 

steam supply to the turbine-driven 

auxiliary feedwater pump (AFW) or the 

inoperability of the turbine-driven AFW 

pump under certain operating mode 

restrictions. 

Date of issuance: February 28, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No.: 235. 

Facility Operating License No NPF–

49: Amendment revised the License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: April 11, 2006 (70 FR 18372). 

The supplements dated August 14, 

2006, and January 2, 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 amendments is contained in a 

Safety Evaluation dated February 28, 

2007. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc. 

Docket Nos. 50–313 and 50–368, 

Arkansas Nuclear One, Units 1 and 2, 

Pope County, Arkansas. 

Date of amendment request: October 

25, 2005, as supplemented by letter 

dated March 20, 2006. 


sroberts on PROD1PC70 with NOTICES
Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 14311 

Brief description of amendments: The 

changes addressed inventory and 

inspection requirements associated with 

the emergency cooling pond, which is a 

common cooling water source for both 

units during conditions that may render 

the normal cooling water source 

(Dardanelle Reservoir) unavailable. 

Date of issuance: March 9, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: Unit 1–229, Unit 

2–271. 

Renewed Facility Operating License 

Nos. DPR–51 and NPF–6: Amendments 

revised the Operating Licenses and 

Technical Specifications. 

Date of initial notice in Federal 

Register: October 24, 2006 (71 FR 

62309). The supplemental letter dated 

March 20, 2006, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 9, 2007. 

No significant hazards consideration 

comments received: No. 

Nine Mile Point Nuclear Station, LLC 

Docket No. 50–220, Nine Mile Point 

Nuclear Station, Unit No. 1, Oswego 

County, New York. 

Date of application for amendment: 

October 19, 2006, as supplemented by 

letter dated January 5, 2007. 

Brief description of amendment: The 

amendment revises the Surveillance 

Requirement (SR) in Technical 

Specification (TS) 4.1.1.c, ‘‘Scram 

Insertion Times,’’ to modify the 

conditions under which scram time 

testing (STT) of control rods is required, 

and to add a requirement to perform 

STT on a defined portion of control 

rods, at a specified frequency, during 

the operating cycle. The amendment 

also revises the SR in TS 4.1.7.c, 

‘‘Minimum Critical Power Ratio 

(MCPR),’’ to add a requirement to 

determine the MCPR operating limits 

following completion of control rod STT 

per TS 4.1.1.c. 

Date of issuance: March 15, 2007. 

Effective date: March 15, 2007. 

Amendment No.: 193. 

Facility Operating License No. DPR–

63: Amendment revised the License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: December 5, 2006 (71 FR 

70562) The supplemental letter dated 

January 5, 2007, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 15, 2007. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC 

Docket No. 50–311, Salem Nuclear 

Generating Station, Unit No. 2, Salem 

County, New Jersey. 

Date of application for amendment: 

April 6, 2006. 

Brief description of amendment: The 

amendment changed the Technical 

Specifications (TSs) to reduce the 

maximum allowable reactor power level 

when two main steam safety valves are 

inoperable. 

Date of issuance: March 7, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to restart from the steam generator 

replacement outage. 

Amendment No.: 259. 

Facility Operating License No. DPR–

75: The amendment revised the TSs and 

the License. 

Date of initial notice in Federal 

Register: November 7, 2006 (71 FR 

65144). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 7, 2007. 

No significant hazards consideration 

comments received: No. 

R.E. Ginna Nuclear Power Plant, LLC 

Docket No. 50–244, R.E. Ginna 

Nuclear Power Plant, Wayne County, 

New York. 

Date of application for amendment: 

May 1, 2006, as supplemented by letter 

dated November 3, 2006. 

Brief description of amendment: The 

amendment revises the steam generator 

tube integrity Technical Specifications 

consistent with the Nuclear Regulatory 

Commission’s approved Technical 

Specification Task Force (TSTF) 

Standard Technical Specification 

Change Traveler, TSTF–449, ‘‘Steam 

Generator Tube Integrity,’’ Revision 4. 

Date of issuance: March 1, 2007. 

Effective date: As of the date of 

issuance to be implemented within 90 

days. 

Amendment No.: 100. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: June 6, 2006 (71 FR 32605). 

The supplemental letter dated 

November 3, 2006, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 1, 2007. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority (TVA) 

Docket No. 50–259, Browns Ferry 

Nuclear Plant (BFN), Unit 1, Limestone 

County, Alabama. 

Date of application for amendment: 

September 22, 2006. 

Brief description of amendment: The 

amendment supplements a June 28, 

2004, request to increase the licensed 

thermal power from 3293 megawatt 

thermal (MWt) to 3952 MWt, an 

approximate 20% increase in thermal 

power. This supplement requests 

interim approval of an increase in 

licensed thermal power from 3293 MWt 

to 3458 MWt with an attendant 30-psi 

increase in reactor pressure. This 

represents an approximate 5% increase 

above the original licensed thermal 

power of 3293 MWt. An interim 

approval would provide for operation at 

105% power until such time as certain 

steam dryer analyses can be completed. 

The NRC staff’s review of the remainder 

of the June 2004 application would 

resume upon receipt of the satisfactorily 

completed steam dryer analyses. 

Date of issuance: March 6, 2007. 

Effective date: Date of issuance, to be 

implemented prior to restart. 

Amendment No.: 269. 

Renewed Facility Operating License 

No. DPR–33: Amendment revised the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: October 10, 2006 (71 FR 

59532). The Commissions related 

evaluation of the amendment is 

contained in an Environmental 

Assessment dated February 12, 2007 (72 

FR 6612), and in a Safety Evaluation 

dated March 6, 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: 

March 28, 2006, as supplemented by 

letter dated November 17, 2006. 

Brief description of amendment: The 

amendment deleted references to 

specific isolation valves in the chemical 


sroberts on PROD1PC70 with NOTICES
14312 Federal Register / Vol. 72, No. 58 / Tuesday, March 27, 2007 / Notices 

and volume control system (CVCS) and 

modified to allow (1) an exception for 

decontamination activities and (2) an 

exception for CVCS resin vessel 

operation. These are changes to TS 

3.3.9, ‘‘Boron Dilution Mitigation 

System (BDMS),’’ and TS 3.9.2, 

‘‘Unborated Water Source Isolation 

Valves.’’

Date of issuance: March 8, 2007. 

Effective date: As of its date of 

issuance and shall be implemented 

within 90 days of the date of issuance. 

Amendment No.: 181. 

Facility Operating License No. NPF–

30: The amendment revised the 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: May 9, 2006 (71 FR 27004). 

The supplemental letter dated 

November 17, 2006, did not expand the 

scope of the application as originally 

noticed, and did not change the NRC 

staff’s original proposed no significant 

hazards consideration determination 

published in the Federal Register. The 

Commission’s related evaluation of the 

amendment is contained in a Safety 

Evaluation dated March 8, 2007. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company 

Docket Nos. 50–338 and 50–339, 

North Anna Power Station, Units 1 and 

2, Louisa County, Virginia. 

Date of application for amendment: 

October 3, 2006, as supplemented by 

letter dated January 24, 2007. 

Brief description of amendment: The 

proposed amendments revised the 

Technical Specifications (TSs) and 

licensing basis to support the resolution 

of the Nuclear Regulatory Commission’s 

(NRC’s) Generic Safety Issue (GSI) 191, 

assessment of debris accumulation on 

containment sump performance and its 

impact on emergency recirculation 

during an accident, and NRC Generic 

Letter (GL) 2004–02. 

Date of issuance: March 13, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 250 and 230. 

Renewed Facility Operating License 

Nos. NPF–4 and NPF–7: Amendments 

change the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: December 5, 2006 (71 FR 

70563). The supplement dated January 

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. The 

Commission’s related evaluation of the 

amendments is contained in a Safety 

Evaluation dated March 13, 2007. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 19th day 

of March 2007. 

For the Nuclear Regulatory Commission. 

John W. Lubinski, 

Acting Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E7–5342 Filed 3–26–07; 8:45 am] 

BILLING CODE 7590–01–P 

POSTAL SERVICE 

Board of Governors; Sunshine Act 

Meeting 

Board Votes to Close March 19, 2007 

Meeting 

At its teleconference meeting on March 

16, 2007, the Board of Governors of the 

United States Postal Service voted 

unanimously to close to public 

observation its meeting scheduled for 

March 19, 2007, in Washington, DC, via 

teleconference. The Board determined 

that prior public notice was not 

possible. 

ITEM CONSIDERED: Postal Regulatory 

Commission Opinion and 

Recommended Decision in Docket No. 

R2006–1, Postal Rate and Fee Changes. 

GENERAL COUNSEL CERTIFICATION: The 

General Counsel of the United States 

Postal Service has certified that the 

meeting was properly closed under the 

Government in the Sunshine Act. 

FOR FURTHER INFORMATION CONTACT: 

Requests for information about the 

meeting should be addressed to the 

Secretary of the Board, Wendy A. 

Hocking, at (202) 268–4800. 

Wendy A. Hocking, 

Secretary. 

[FR Doc. 07–1487 Filed 3–22–07; 4:43 pm] 

BILLING CODE 7710–12–M 

POSTAL SERVICE 

Board of Governors; Sunshine Act 

Meeting 

TIME AND DATE: 8 a.m., Wednesday, 

March 28, 2007. 

PLACE: Washington, DC, at U.S. Postal 

Service Headquarters, 475 L’Enfant 

Plaza, SW. 

STATUS: Closed. 

MATTERS TO BE CONSIDERED: 

Wednesday, March 28 at 8 a.m. 

(Closed) 

1. Strategic Issues. 

2. Rates Implementation. 

3. Labor Negotiations Update. 

4. Financial Update. 

5. Personnel Matters and 

Compensation Issues. 

6. Governors’ Executive Session—

Discussion of prior agenda items and 

Board Governance. 

FOR FURTHER INFORMATION CONTACT: 

Wendy A. Hocking, Secretary of the 

Board, U.S. Postal Service, 475 L’Enfant 

Plaza, SW., Washington, DC 20260–

1000. Telephone (202) 268–4800. 

Wendy A. Hocking, 

Secretary. 

[FR Doc. 07–1488 Filed 3–22–07; 4:43 pm] 

BILLING CODE 7710–12–M 

POSTAL SERVICE 

Board of Governors; Sunshine Act 

Meeting 

Board Votes to Close March 16, 2007 

Meeting 

At its teleconference meeting on 

March 14, 2007, the Board of Governors 

of the United States Postal Service voted 

unanimously to close to public 

observation its meeting scheduled for 

March 16, 2007, in Washington, DC, via 

teleconference. The Board determined 

that prior public notice was not 

possible. 

ITEM CONSIDERED: Postal Regulatory 

Commission Opinion and 

Recommended Decision in Docket No. 

R2006–1, Postal Rate and Fee Charges. 

GENERAL COUNSEL CERTIFICATION: The 

General Counsel of the United States 

Postal Services has certified that the 

meeting was properly closed under the 

Government in the Sunshine Act. 

FOR FURTHER INFORMATION CONTACT: 

Requests for information about the 

meeting should be addressed to the 

Secretary of the Board, Wendy A. 

Hocking, at (202) 268–4800. 

Wendy A. Hocking, 

Secretary. 

[FR Doc. 07–1489 Filed 3–22–07; 4:43 pm] 

BILLING CODE 7710–12–M 


sroberts on PROD1PC70 with NOTICES