5078 Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 

home page site for 60 days after the 

signature date of this notice. 

Comments and questions about the 

information collection requirements 

may be directed to the NRC Clearance 

Officer, Brenda Jo. Shelton (T–5 F53), 

U.S. Nuclear Regulatory Commission, 

Washington, DC 20555–0001, by 

telephone at 301–415–7233, or by 

Internet electronic mail to 

infocollects@nrc.gov. 

Dated at Rockville, Maryland, this 24th of 

January 2006. 

For the Nuclear Regulatory Commission. 

Brenda Jo. Shelton, 

NRC Clearance Officer, Office of Information 

Services. 

[FR Doc. E6–1293 Filed 1–30–06; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Sunshine Act Meeting 

DATES: Weeks of January 30, February 6, 

13, 20, 27, March 6, 2006. 

PLACE: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland. 

STATUS: Public and Closed. 

MATTERS TO BE CONSIDERED: 

Week of January 30, 2006 

Tuesday, January 31, 2006 

9:25 a.m. Affirmation Session (Public 

Meeting). 

a. FIRSTENERGY Nuclear Operating 

Co. (Beaver Valley Power Station, 

Unit Nos. 1 & 2; Davis Besse Power 

Station, Unit 1; Perry Nuclear 

Power Plant, Unit No. 1), Docket 

Nos. 50–334–LT, 50–346–LT, 50–

412–LT, & 50–440–LT. 

b. Private Fuel Storage (Independent 

Spent Fuel Storage installation) 

Docket No. 72–22–ISFSI. 

c. Motion to Reopen the Millstone 

License Renewal Proceedings Filed 

by Connecticut Coalition Against 

Millstone. 

9:30 a.m. Briefing on Strategic 

Workforce Planning and Human 

Capital Initiatives (Public Meeting). 

(Contact: Kristen Davis, 301–415–

7108.) 

This meeting will be webcast live at 

the Web address http://www.nrc.gov. 

Wednesday, February 1, 2006 

9:30 a.m. Discussion of Security Issues 

(Closed—Ex. 1 & 3). 

Thursday, February 2, 2006 

1:30 p.m. Briefing on Sensitive 

Unclassified Non-Safeguards 

Information (SUNSI) Policy (Public 

Session and Closed Session—Ex. 2). 

(Contact: Edward Baker, 301–415–

8700.) 

Open portion of this meeting will be 

webcast live at the Web address 

http://www.nrc.gov. 

Week of February 6, 2006—Tentative 

Monday, February 6, 2006 

9:30 a.m. Briefing on Materials 

Degradation Issues and Fuel 

Reliability (Public Meeting). 

(Contact: Jennifer Uhle, 301–415–

6200.) 

This meeting will be webcast live at 

the Web address http://www.nrc.gov 

2 p.m. Discussion of Security Issues 

(Closed—Ex. 1). 

Wednesday, February 8, 2006 

9:30 a.m. Briefing on Office of Nuclear 

Materials Safety and Safeguards 

(NMSS) Programs, Performance, 

and Plans—Materials Safety (Public 

Meeting). (Contact: Teresa Mixon, 

301–415–7474; Derek Widmayer, 

301–415–6677.) 

This meeting will be webcast live at 

the Web address http://www.nrc.gov. 

1:30 p.m. Briefing on Office of Research 

(RES) Programs, Performance and 

Plans (Public Meeting). (Contact: 

Gene Carpenter, 301–415–7333.) 

This meeting will be webcast live at 

the Web address http://www.nrc.gov. 

Week of February 13, 2006—Tentative 

Tuesday, February 14, 2006 

2 p.m. Briefing on Office of Nuclear 

Materials Safety and Safeguards 

(NMSS) Programs, Performance, 

and Plans—Waste Safety (Public 

Meeting). (Contact: Teresa Mixon, 

301–415–7474; Derek Widmayer, 

301–415–6677.) 

The meeting will be webcast live at 

the Web address http://www.nrc.gov. 

Wednesday, February 15, 2006 

9:30 a.m. Briefing on Office of Chief 

Financial Officer (CFO) Programs, 

Performance, and Plans (Public 

Meeting). (Contact: Edward New, 

301–415–5646.) 

This meeting will be webcast live at 

the Web address http://www.nrc.gov. 

Week of February 20, 2006—Tentative 

There are no meetings scheduled for 

the Week of February 20, 2006. 

Week of February 27, 2006—Tentative 

There are no meetings scheduled for 

the Week of February 27, 2006. 

Week of March 6, 2006—Tentative 

There are no meetings scheduled for 

the Week of March 6, 2006. 

*The schedule for Commission 

meetings is subject to change on short 

notice. To verify the status of meetings 

call (recording)—(301) 415–1292. 

Contact person for more information: 

Michelle Schroll, (301) 415–1662. 

* * * * * 

The NRC Commission Meeting 

Schedule can be found on the Internet 

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

policy-making/schedule.html. 

* * * * * 

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g., 

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

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

301–415–2100, or by e-mail at 

aks@nrc.gov. Determinations on 

requests for reasonable accommodation 

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

* * * * * 

This notice is distributed by mail to 

several hundred subscribers; if you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301) 415–1969. 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to dkw@nrc.gov. 

Dated: January 26, 2006. 

R. Michelle Schroll, 

Office of the Secretary. 

[FR Doc. 06–925 Filed 1–27–06; 11:26 am] 

BILLING CODE 7590–01–M 

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a.(2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 


hsrobinson on PROD1PC70 with NOTICES
Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 5079 

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 January 6, 

2006 to January 19, 2006. The last 

biweekly notice was published on 

January 17, 2006 (71 FR 2586). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. Within 60 days after the 

date of publication of this notice, the 

licensee may file a request for a hearing 

with respect to issuance of the 

amendment to the subject facility 

operating license and any person whose 

interest may be affected by this 

proceeding and who wishes to 

participate as a party in the proceeding 

must file a written request for a hearing 

and a petition for leave to intervene. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules and 

Directives Branch, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be delivered to Room 6D22, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

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

Copies of written comments received 

may be examined at the Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area 01F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. The filing of 

requests for a hearing and petitions for 

leave to intervene is discussed below. 

Within 60 days after the date of 

publication of this notice, the licensee 

may file a request for a hearing with 

respect to issuance of the amendment to 

the subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

intervene. Requests for a hearing and a 

petition for leave to intervene shall be 

filed in accordance with the 

Commission’s ‘‘Rules of Practice for 

Domestic Licensing Proceedings’’ in 10 

CFR part 2. Interested persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

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 


hsrobinson on PROD1PC70 with NOTICES
5080 Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 

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. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units 1, 2, and 3, 

Maricopa County, Arizona 

Date of amendments request: 

December 23, 2005. 

Description of amendments request: 

The amendments would increase the 

emergency diesel generator (EDG) 

allowed out of service time (AOT) from 

72 hours to 10 days, allow EDG starting 

air receiver pressure to momentarily 

drop below limits during successful 

starting of an EDG, and remove from the 

Technical Specifications the statement 

that the two groups of pressurizer 

heaters are capable of being powered 

from an emergency power supply. 

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 Technical Specification (TS) 

change to increase the emergency diesel 

generator (EDG) allowed out of service time 

(AOT) from 72 hours to 10 days will not 

cause an accident to occur and will not result 

in any change in the operation of the 

associated accident mitigation equipment. 

The EDGs are not accident initiators. The 

EDGs are designed to mitigate the 

consequences of previously evaluated 

accidents including a loss of offsite power. 

Extending the AOT for a single EDG would 

not affect the previously evaluated accidents 

since the remaining EDG supporting the 

redundant Engineered Safety Features (ESF) 

systems would continue to be available to 

perform the accident mitigation functions. 

The duration of this TS AOT considers that 

there is a minimal possibility that an 

accident will occur while a component is 

removed from service. A risk informed 

assessment was performed which concluded 

that the increase in plant risk is small and 

consistent with the guidance contained in 

Regulatory Guide 1.177, ‘‘An Approach for 

Plant-Specific, Risk-Informed 

Decisionmaking: Technical Specifications.’’

The design basis accidents will remain the 

same postulated events described in the 

PVNGS [Palo Verde Nuclear Generating 

Station] Updated Final Safety Analysis 

Report (UFSAR). In addition, extending the 

EDG AOT will not impact the consequences 

of an accident previously evaluated. The 

consequences of previously evaluated 

accidents will remain the same during the 

proposed 10 day AOT as during the current 

72 hour AOT. The ability of the remaining 

TS-required EDG to mitigate the 

consequences of an accident will not be 

affected since no additional failures are 

postulated while equipment is inoperable 

within the TS AOT. The remaining EDG is 

sufficient to mitigate the consequences of any 

design basis accident. 

The proposed addition of a note to 

Condition F of TS 3.8.3, would allow EDG 

starting air receiver pressure to momentarily 

drop below limits during successful starting 

of an EDG. The EDG air starting system will 

not be operated or be configured any 

differently than that which it is currently 

required and designed for. This proposed 

change will only add a note for clarification 

to Condition F of TS 3.8.3. This note 

describes entering this Condition is not 

necessary when the EDG starts normally and 

is operating per required procedures. 

Momentary transients outside the air receiver 

pressure range do not invalidate the 

successful start and running of the EDG. A 

successful start of the EDG indicates the 

starting air system has performed its required 

safety function. This proposed change will 

not increase the probability or consequence 

of an accident previously evaluated. 

The proposed TS change associated with 

the requirements for the pressurizer heaters 

to be supplied by emergency power will not 

result in any change in plant design. These 

components will continue to be powered 

from Class 1E power sources as described in 

the proposed TS Bases change associated 

with this change. As a result, the operation 

and reliability of the pressurizer heaters will 

not be affected by the proposed description 

change. In addition, operation of the 

pressurizer heaters is not assumed to mitigate 

any design basis accident. The proposed 

changes will not cause an accident to occur 

and will not result in a change in the 

operation of any accident mitigation 

equipment. The design basis accidents 

remain the same postulated events described 

in the PVNGS UFSAR. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different [kind of] 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve a 

change in the design, configuration, or 

method of operation of the plant that could 

create the possibility of a new or different 

[kind of] accident. Equipment will be 

operated in the same configuration and 

manner that is currently allowed and 

designed for. The proposed changes do not 

introduce any new failure modes. This 

license amendment request does not impact 


hsrobinson on PROD1PC70 with NOTICES
Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 5081 

any plant systems that are accident initiators 

or adversely impact any accident mitigating 

systems. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

[kind of] accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The EDG reliability and availability are 

monitored and evaluated, in accordance with 

10 CFR 50.65 (Maintenance Rule) 

performance criteria, to assure EDG out of 

service times do not degrade operational 

safety over time. Extension of the EDG AOT 

will not erode the reduction in severe 

accident risk that was achieved with 

implementation of the Station Blackout 

(SBO) rule (10 CFR 50.63) or affect any safety 

analyses assumptions or inputs. The SBO 

coping analysis is unaffected by the AOT 

extension since the EDGs are not assumed to 

be available during the coping period. The 

assumptions used in the coping analysis 

regarding EDG reliability are unaffected since 

preventive maintenance and testing will 

continue to be performed to maintain the 

reliability assumptions. 

Accident mitigation functions will be 

maintained by the other TS-required EDG 

availability to supply power to the safety 

related Class 1E electrical loads. The 

availability of the TS-required offsite power, 

combined with the availability of the PVNGS 

SBO Gas Turbine Generators (GTGs) and the 

use of the Configuration Risk Management 

Program required by 10 CFR 50.65(a)(4), 

provide adequate compensation for the small 

incremental increase in plant risk of the 

proposed EDG AOT extension. This small 

increase in plant risk while operating is offset 

by a reduction in shutdown risk resulting 

from the increased availability and reliability 

of the EDGs during refueling outages, and 

avoiding transition risk incurred during 

unplanned plant shutdowns. In addition, the 

calculated risk measures associated with the 

proposed AOT are below the acceptance 

criteria defined in Regulatory Guide 1.177. 

The proposed change to add a note to 

Condition F of TS 3.8.3 does not involve 

changes to setpoints or limits established or 

assumed by the accident analyses. This note 

only applies to those occasions when after a 

successful start of an EDG has occurred and 

the starting air receiver pressure has 

momentarily dropped below its limit. This 

change allows for not declaring the EDG 

inoperable solely due to this momentary drop 

in pressure during a successful start of the 

EDG. No safety margin will be impacted by 

this change. 

The proposed TS change associated with 

the wording description of LCO [Limiting 

Condition of Operation] 3.4.9, ‘‘Pressurizer,’’

for the requirement of the pressurizer heaters 

to be supplied by emergency power does not 

adversely affect equipment design or 

operation, and there are no changes being 

made to the TS-required safety limits or 

system settings that would adversely affect 

plant safety. The emergency power 

requirements for the pressurizer heaters, 

which came from the Three Mile Island 

(TMI) action item requirement II.E.3.1, 

‘‘Emergency Power Requirements for 

Pressurizer Heater,’’ of NUREG–0737, 

‘‘Clarification of TMI Action Plan 

Requirements,’’ will continue to be met. The 

pressurizer heaters used to satisfy the 

NUREG–0737 and LCO 3.4.9 requirements 

are, by design, permanently connected to 

Class 1E power supplies as described in the 

PVNGS Updated Final Safety Analyses 

Report, Section 18.II.E.3.1. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on that 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the request 

for amendments involves no significant 

hazards consideration. 

Attorney for licensee: Kenneth C. 

Manne, Senior Attorney, Arizona Public 

Service Company, P.O. Box 52034, Mail 

Station 7636, Phoenix, Arizona 85072–

2034. 

NRC Branch Chief: David Terao. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit 1, Lake 

County, Ohio 

Date of amendment request: 

November 15, 2005. 

Description of amendment request: 

The proposed change modifies the 

technical specifications (TS) to clarify 

the wording of emergency closed 

cooling water (ECCW) Surveillance 

Requirement (SR) 3.7.10.2. The current 

wording in SR 3.7.10.2 requires that 

automatic valves on the ECCW system 

actuate on an actuation signal. However, 

the TS Bases for the SR identify more 

than just valves tested to include the 

automatic start capability of the ECCW 

pump in each subsystem. Therefore, the 

wording of this SR would be modified 

to clarify that its purpose is to verify 

actuation of the entire subsystem on an 

actual or simulated signal, rather than 

just verify valve actuation. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below: 

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

There are no physical modifications being 

made to any plant system or component. The 

only change is to a Surveillance Requirement 

within the Technical Specifications, in order 

to improve understanding and avoid 

misinterpretation of the requirements. The 

original intent of ECCW SR 3.7.10.2 is 

maintained by the change being proposed. 

The revised Technical Specification 

requirements do not impact initiators of 

previously evaluated accidents or transients. 

The specification being revised is 

associated with a system used to mitigate the 

consequences of accidents. The change to the 

wording of ECCW SR 3.7.10.2 does not 

impact the capability of the associated 

system to perform its required function. The 

reworded ECCW SR more clearly requires 

that the system[’]s total actuation capability 

be maintained. 

The change does not affect how plant 

systems are controlled or operated or tested. 

The change continues to provide 

confirmation of the capability of plant 

components to respond as required to 

mitigate the consequences of events. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

There are no physical modifications being 

made to any plant system or component, and 

the proposed change introduces no new 

method of operation of the plant, or its 

systems or components. Therefore, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. The proposed change does not involve 

a significant reduction in a margin of safety. 

The change to the ECCW SR continues to 

ensure the ECCW subsystems are tested on 

the same periodicity to verify their capability 

to respond to actuation signals from the 

Emergency Core Cooling System (ECCS) 

Instrumentation Functions of Low Water 

Level and High Drywell Pressure. Therefore, 

the necessary function of the Technical 

Specification requirements is maintained, 

and the proposed change does not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David W. 

Jenkins, Attorney, FirstEnergy 

Corporation, Mail Stop A–GHE–107, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Mindy Landau, 

Acting. 

Nuclear Management Company, LLC, 

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

Island Nuclear Generating Plant, Units 

1 and 2, Goodhue County, Minnesota 

Date of amendment request: 

December 13, 2005. 

Description of amendment request: 

The proposed amendments would 

revise technical specification (TS) 

requirements for surveillance 


hsrobinson on PROD1PC70 with NOTICES
5082 Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 

requirements for containment integrated 

leakage rate testing in TS 5.5.14.a to 

allow a one-time extension of the 

interval between reactor containment 

vessel integrated leakage rate tests 

(ILRTs) from 10 to 15 years. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This license amendment proposes to revise 

the Technical Specifications to allow for the 

one time extension of the containment 

integrated leakage rate test interval from 10 

to 15 years. The containment vessel function 

is purely mitigative. There are no design 

basis accidents initiated by a failure of the 

containment leakage mitigation function. The 

extension of the containment integrated 

leakage rate test interval will not create any 

adverse interactions with other systems that 

could result in initiation of a design basis 

accident. Therefore, the probability of 

occurrence of an accident previously 

evaluated is not significantly increased. 

The potential consequences of the 

proposed change have been quantified by 

analyzing the changes in risk that would 

result from extending the containment 

integrated leakage rate test interval from 10 

to 15 years. The increase in risk in terms of 

person-rem per year within 50 miles 

resulting from design basis accidents was 

estimated to be of a magnitude that NUREG–

1493, ‘‘Performance-Based Containment 

Leak-Test Program’’, indicates is 

imperceptible. The Nuclear Management 

Company has also analyzed the increase in 

risk in terms of the frequency of large early 

releases from accidents. The increase in the 

large early release frequency resulting from 

the proposed extension was determined to be 

within the guidelines published in 

Regulatory Guide 1.174, ‘‘An Approach for 

Using Probabilistic Risk Assessment in RiskInformed Decisions on Plant-Specific 

Changes to the Current Licensing Basis’’. 

Additionally, the proposed change maintains 

defense-in-depth by preserving a reasonable 

balance among prevention of core damage, 

prevention of containment failure, and 

consequence mitigation. The Nuclear 

Management Company has determined that 

the increase in conditional containment 

failure probability from reducing the 

containment integrated leakage rate test 

frequency from 1 test per 10 years to 1 test 

per 15 years would be small. 

Continued containment integrity is also 

assured by the history of successful 

containment integrated leakage rate tests, and 

the established programs for local leakage 

rate testing and in-service inspections which 

are unaffected by the proposed change. 

Therefore, the probability of occurrence or 

the consequences of an accident previously 

analyzed are not significantly increased. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change to extend the 

containment integrated leakage rate test 

interval from 10 to 15 years does not create 

any new or different accident initiators or 

precursors. The length of the containment 

integrated leakage rate test interval does not 

affect the manner in which any accident 

begins. The proposed change does not create 

any new failure modes for the containment 

and does not affect the interaction between 

the containment and any other system. Thus, 

the proposed changes do not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The risk-based margins of safety associated 

with the containment integrated leakage rate 

test are those associated with the estimated 

person-rem per year, the large early release 

frequency, and the conditional containment 

failure probability. The Nuclear Management 

Company has quantified the potential effect 

of the proposed change on these parameters 

and determined that the effect is not 

significant. The non-risk-based margins of 

safety associated with the containment 

integrated leakage rate test are those involved 

with its structural integrity and leak 

tightness. The proposed change to extend the 

containment integrated leakage rate test 

interval from 10 to 15 years does not 

adversely affect either of these attributes. The 

proposed change only affects the frequency at 

which these attributes are verified. Therefore, 

the proposed change does not involve a 

significant reduction in margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: Jonathan Rogoff, 

Esquire, Vice President, Counsel & 

Secretary, Nuclear Management 

Company, LLC, 700 First Street, 

Hudson, WI 54016. 

NRC Acting Branch Chief: Timothy 

Kobetz. 

PPL Susquehanna, LLC, Docket Nos. 50–

387 and 50–388, Susquehanna Steam 

Electric Station, Units 1 and 2 (SSES 1 

and 2), Luzerne County, Pennsylvania 

Date of amendment request: 

November 18, 2005. 

Description of amendment request: 

The proposed amendment would 

change the SSES 1 and 2 Technical 

Specifications (TSs) to implement the 

Average Power Range Monitor/Rod 

Block Monitor/Technical 

Specifications/Maximum Extended 

Load Line Limit Analysis (ARTS/ 

MELLLA). Specifically, the average 

power range monitor (APRM) flowbiased scram and rod block trip 

setpoints would be revised to permit 

operation in the MELLLA region. The 

current flow-biased rod block monitor 

(RBM) would also be replaced by a 

power dependent RBM implemented 

through the referenced proposed 

upgrade to a digital power range 

neutron monitor system (PRNMS). The 

change from the flow-biased RBM to the 

power-dependent RBM would also 

require new trip setpoints. In addition, 

the flow-biased APRM scram and rod 

block trip setdown requirement would 

be replaced by more direct power and 

flow-dependent thermal limits to reduce 

the need for APRM gain adjustments, 

and to allow more direct thermal limits 

administration during operation other 

than rated conditions. Finally, the 

proposed amendment would change the 

methods used to evaluate the annulus 

pressurization (AP), mass blowdown, 

and early release resulting from the 

postulated recirculation suction line 

break (RSLB). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

Proposed Change No. 1: The proposed 

change eliminates the Average Power Range 

Monitor (APRM) flow-biased scram and rod 

block trip setpoint setdown requirements 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 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. Therefore, the proposed change 

will not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

Proposed Change No. 2: The proposed 

change expands the power and flow 

operating domain by relaxing the restrictions 

imposed by the formulation of the APRM 

flow-biased scram and rod block trip 

setpoints and the replacement of the current 

flow-biased RBM with a new power 

dependent RBM, which will be implemented 

using a digital Power Range Neutron 

Monitoring System (PRNMS). The APRM and 

RBM are not involved in the initiation of any 


hsrobinson on PROD1PC70 with NOTICES
Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 5083 

accident; and the APRM flow-biased scram 

and rod block functions are not credited in 

any PPL safety licensing analyses. 

The analysis of the instrument line break 

event resulted in an insignificant change in 

the radiological consequences. The change 

for the instrument line break was an 

insignificant increase of 0.1 Rem. 

Since the proposed changes will not affect 

any accident initiator, or introduce and 

initial conditions that would result in NRC 

approved criteria being exceeded, and since 

the APRM and RBM will remain capable of 

performing their design functions, the 

proposed change will not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Proposed Change No. 3: The methods used 

to evaluate Annulus Pressurization (AP) and 

mass blowdown and energy releases resulting 

from the postulated Recirculation Suction 

Line Break (RSLB) at the MELLLA conditions 

are changed to use more realistic, but still 

conservative, methods of analysis to 

determine an AP mass and energy release 

profile for AP loads resulting from the 

postulated RSLB. The releases resulting from 

the RSLB at off-rated conditions have been 

demonstrated to be bounded by the current 

design basis loads. Since the proposed 

changes do not affect any accident initiator 

and since the RSLB AP releases remain 

bounded by the current design basis, the 

proposed changes do not involve a 

significant increase in the probability or 

radiological consequences of an accident 

previously evaluated. Therefore the proposed 

changes do not involve a significant increase 

in the probability or consequences of any 

accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Proposed Change No. 1: The proposed 

change eliminates the Average Power Range 

Monitor (APRM) flow-biased scram and rod 

block setpoint setdown requirements and 

substitutes power and flow dependent 

adjustments to the Minimum Critical Power 

Ratio (MCPR) and Linear Heat Generation 

Rate (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. Therefore, the 

proposed change will not create the 

possibility of a new or different kind of 

accident previously evaluated. 

Proposed Change No. 2: The proposed 

change expands the power and flow 

operating domain by relaxing the restrictions 

imposed by the formulation of the APRM 

flow-biased scram and rod block trip 

setpoints and the replacement of the current 

flow-biased RBM with a new power 

dependent RBM, which will be implemented 

using a digital Power Range Neutron 

Monitoring System (PRNMS). Changing the 

formulation for the APRM flow-biased scram 

and rod block trip setpoints and from a flowbiased RBM to a power dependent RBM does 

not change their respective functions and 

manner of operation. The change does not 

introduce a sequence of events or introduce 

a new failure mode that would create a new 

or different type of accident. The APRM 

flow-biased rod block trip setpoint will 

continue to block control rod withdrawal 

when core power significantly exceeds 

normal limits and approaches the scram 

level. The APRM flow-biased scram trip 

setpoint will continue to initiate a scram if 

the increasing power/flow condition 

continue beyond the APRM flow-biased rod 

block setpoint. 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 changes. 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. 

Proposed Change No. 3: The methods used 

to evaluate Annulus Pressurization (AP) and 

mass blowdown and energy releases resulting 

from the postulated Recirculation Suction 

Line Break (RSLB) at the MELLLA conditions 

are changed to use more realistic, but still 

conservative, methods of analysis to 

determine an AP mass and energy release 

profile for AP loads resulting from the 

postulated RSLB. The proposed changes to 

the methods of analysis to determine AP 

mass and energy releases resulting from the 

postulated RSLB do not change the design 

function or operation of any plant 

equipment. No new failure mechanisms, 

malfunctions, or accident initiators are being 

introduced by the proposed changes. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

Proposed Change No. 1: The proposed 

change eliminates the Average Power Range 

Monitor (APRM) flow-biased scram and rod 

block setpoint setdown requirements and 

substitutes power and flow dependent 

adjustments to the Minimum Critical Power 

Ratio (MCPR) and Linear Heat Generation 

Rate (LHGR) thermal limits. Replacement of 

the APRM setpoint setdown requirement 

with power and flow dependent adjustments 

to the MPR and LHGR thermal limits will 

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 

MPR 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 10 CFR 50.46 acceptance 

criteria for the performance of the Emergency 

Core Cooling System (ECCS) following 

postulated Loss-Of-Coolant Accidents 

(LOCAs) will continue to be met. Therefore, 

the proposed change will not involve a 

significant reduction in a margin of safety. 

Proposed Change No. 2: The proposed 

change expands the power and flow 

operating domain by relaxing the restrictions 

imposed by the formulation of the APRM 

flow-biased scram and rod block trip 

setpoints and the replacement of the current 

flow-biased RBM with a new power 

dependent RBM, which will be implemented 

using a digital Power Range Neutron 

Monitoring System (PRNMS). The APRM 

flow-biased rod block trip setpoint will 

continue to block control rod withdrawal 

when core power significantly exceeds 

normal limits and approaches the scram 

level. The APRM flow-biased scram trip 

setpoint will continue to initiate a scram if 

the increasing power/flow condition 

continues beyond the APRM flow-biased rod 

block setpoint. The RBM will continue to 

prevent rod withdrawal when the power 

dependent RBM rod block setpoint is 

reached. The MPR and LHGR thermal limits 

will be developed to ensure that fuel thermal 

mechanical design bases shall remain within 

the licensing limits during a rod withdrawal 

error event and to ensure that the MPR safety 

limit will not be violated as a result of a 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. Anticipated 

operational occurrences and postulated 

accident within the expanded operating 

domain will be evaluated using NRC 

approved methods. Therefore, the proposed 

change will not involve a significant 

reduction in the margin of safety. 

Proposed Change No. 3: The methods used 

to evaluate Annulus Pressurization (AP) and 

mass blowdown and energy releases resulting 

from the postulated Recirculation Suction 

Line Break (RSLB) at the MELLLA conditions 

are changed to use more realistic, but still 

conservative, methods of analysis to 

determine an AP mass and energy release 

profile for AP loads resulting from the 

postulated RSLB. Mass and energy releases 

for AP loads resulting from the postulated 

RSLB remain bounded by the current design 

basis releases. 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: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179. 

NRC Branch Chief: Richard J. Laufer. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of amendment request: October 

11, 2005. 


hsrobinson on PROD1PC70 with NOTICES
5084 Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 

Description of amendment request: 

The proposed amendment would 

remove the Technical Specification (TS) 

3.1.5 requirement for the Standby 

Liquid Control (SLC) system to be 

operable in Operational Condition 5 

(refueling) with any control rod 

withdrawn. Corresponding changes 

would also be made to the SLC 

Initiation sections of Tables 3.3.2–1 and 

4.3.2–1. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes to delete the 

operability requirement for the SLC System 

in OPERATIONAL CONDITION 5* 

(OPERATIONAL CONDITION 5 with any 

control rod withdrawn) does not affect the 

probability or consequences of an accident 

previously evaluated. In STARTUP and 

POWER OPERATION, the SLC System is 

required to provide shutdown capability. In 

HOT SHUTDOWN and COLD SHUTDOWN, 

control rods are not able to be withdrawn 

since the reactor mode switch is in 

Shutdown and a control rod block is applied. 

This provides adequate controls to ensure 

that the reactor remains subcritical. Design 

basis accident mitigation scenarios for 

OPERATIONAL CONDITION 5 do not 

depend on, or require, SLC System 

operability. In REFUELING mode, only a 

single control rod can be withdrawn from a 

core cell containing fuel assemblies. 

Demonstration of adequate shutdown margin 

in accordance with TS LIMITING 

CONDITION FOR OPERATION 3.1.1 ensures 

that the reactor will not become critical. 

Since the purpose of the SLC System is to 

bring the reactor to a cold shutdown 

condition from normal power operations and 

maintain it in a cold shutdown condition, 

there is no design basis for the SLC System 

to be required to be OPERABLE when only 

a single control rod can be withdrawn. In 

addition, the reactor protection system and 

the control rod system would continue to be 

able to provide protection in the unlikely 

event that an inadvertent criticality occurs. 

Therefore, these changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated in the UFSAR [updated final safety 

analysis report]. No new accident scenarios, 

failure mechanisms, or limiting single 

failures are introduced as a result of the 

proposed changes. Specifically, no new 

hardware is being added to the plant as part 

of the proposed change, no existing 

equipment is being modified, and no 

significant changes in operations are being 

introduced. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes will not alter any 

assumptions, initial conditions, or results of 

any accident analyses. The purpose of the 

SLC System is to bring the reactor to and 

maintain it in a cold shutdown condition 

following a failure to scram during plant 

operations. The SLC System is not designed 

to terminate an inadvertent criticality during 

REFUELING. Shutdown margin, either 

demonstrated or analytically determined, in 

accordance with Technical Specifications 

and procedural controls, will assure that an 

inadvertent criticality event will not occur 

during REFUELING. In addition, the reactor 

protection system and control rod system 

provide protection in the unlikely event that 

an inadvertent criticality occurs. The 

proposed change does not affect the ability of 

the SLC System to achieve plant shutdown 

under analyzed conditions (POWER 

OPERATION and STARTUP). 

Therefore, this change does not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jeffrie J. Keenan, 

Esquire, Nuclear Business Unit—N21, 

P.O. Box 236, Hancocks Bridge, NJ 

08038. 

NRC Branch Chief: Darrell J. Roberts. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

Systems (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

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

2), Beaver County, Pennsylvania 

Date of application for amendments: 

October 5, 2004, as supplemented 

March 22, August 29, and October 31, 

2005. 

Brief description of amendments: The 

amendments revised the BVPS–1 and 2 

Technical Specifications (TSs) 3/4.3.1, 

‘‘Reactor Trip System Instrumentation,’’

and 3/4.3.2, ‘‘Engineered Safety Feature 

Actuation Instrumentation,’’ to modify 

steam generator (SG) level allowable 

value (AV) setpoints. Specifically, the 

TS changes increased the AVs of the SG 

water level-low-low setpoints from 14.6 

percent and 16 percent to 19.6 percent 

and 20 percent of the narrow range (NR) 

instrument span for BVPS–1 and 2, 

respectively. These are the AVs of 

setpoints specified in TS Table 3.3–1 to 

initiate a reactor trip, and the actuation 

setpoints specified in TS Table 3.3–3 to 


hsrobinson on PROD1PC70 with NOTICES
Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 5085 

start the auxiliary feedwater pumps. 

Also, for BVPS–2, the AV of the SG 

water level-high-high setpoint increased 

from 81.1 percent to 92.7 percent of the 

NR span. This is the AV of a setpoint 

for actuation of the turbine trip and the 

feedwater system isolation specified in 

TS Table 3.3–3. 

Date of issuance: January 11, 2006. 

Effective date: Upon issuance and 

shall be implemented within 60 days. 

Amendment Nos.: 270 and 152. 

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: November 23, 2004 (69 FR 

68183). The supplements dated March 

22, August 29, and October 31, 2005, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the Nuclear 

Regulatory Commission 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 January 11, 

2006. 

No significant hazards consideration 

comments received: No. 

FPL Energy Seabrook, LLC, Docket No. 

50–443, Seabrook Station, Unit No. 1, 

Rockingham County, New Hampshire 

Date of amendment request: January 

10, 2005. 

Description of amendment request: 

The amendment revised the Seabrook 

Station, Unit No. 1, Technical 

Specifications (TSs) to extend the 

interval for the performance of 

Containment Air Lock Interlock 

Surveillance Requirement 4.6.1.3 from 6 

months to 24 months. 

Date of issuance: January 6, 2006. 

Effective date: As of its date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 106. 

Facility Operating License No. NPF–

86: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: May 24, 2005 (70 FR 29796). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated January 6, 2006. 

No significant hazards consideration 

comments received: No. 

Florida Power and Light Company, et 

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

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

County, Florida 

Date of application for amendments: 

September 1, 2005. 

Brief description of amendments: The 

amendments delete the Technical 

Specification requirements for 

Occupational Radiation Exposure 

Reports and Monthly Operating Reports. 

Date of Issuance: January 13, 2006. 

Effective Date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: 198 and 141. 

Renewed Facility Operating License 

Nos. DPR–67 and NPF–16: Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: October 25, 2005 (70 FR 

61661). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

January 13, 2006. 

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: 

July 21, 2005. 

Brief description of amendments: The 

amendments delete the Technical 

Specification requirements for 

Occupational Radiation Exposure 

Reports and Monthly Operating Reports. 

Date of issuance: January 13, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos: 228 and 224. 

Renewed Facility Operating License 

Nos. DPR–31 and DPR–41: Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: October 25, 2005 (70 FR 

61660). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated January 13, 

2006. 

No significant hazards consideration 

comments received: No. 

Indiana Michigan Power Company, 

Docket Nos. 50–315 and 50–316, Donald 

C. Cook Nuclear Plant, Units 1 and 2, 

Berrien County, Michigan 

Date of application for amendments: 

July 29, 2005. 

Brief description of amendments: The 

amendments revise the units’ Technical 

Specifications by eliminating the 

requirements to submit monthly 

operating reports and occupational 

radiation exposure reports. 

Date of issuance: January 12, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days. 

Amendment Nos.: 292, 274. 

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: December 6, 2005 (70 FR 

72673). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

January 12, 2006. 

No significant hazards consideration 

comments received: No. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: March 8, 

2005, as supplemented by letter dated 

August 18, 2005. 

Brief description of amendment: The 

amendment revised the Technical 

Specification 2.1.1.2 for the single 

recirculation loop Safety Limit 

Minimum Critical Power Ratio value to 

reflect results of a cycle-specific 

calculation. 

Date of issuance: January 4, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 215. 

Facility Operating License No. DPR–

46: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: March 29, 2005 (70 FR 

15944). The supplement dated August 

18, 2005, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 4, 2006. 

No significant hazards consideration 

comments received: No. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: July 21, 

2005. 

Brief description of amendment: The 

amendment revises the technical 

specifications testing frequency for the 

surveillance requirement (SR) in TS 

3.1.4, ‘‘Control Rod Scram Times.’’

Specifically, the proposed change 

would revise the frequency for SR 

3.1.4.2, control rod scram time testing, 

from ‘‘120 days cumulative operation in 

MODE 1’’ to ‘‘200 days cumulative 

operation in MODE 1.’’

Date of issuance: January 5, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 216. 


hsrobinson on PROD1PC70 with NOTICES
5086 Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 

Facility Operating License No. DPR–

46: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 25, 2005 (70 FR 

61661). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

January 5, 2006. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant, Wright County, 

Minnesota 

Date of application for amendment: 

June 30, 2004. 

Brief description of amendment: The 

amendment revised Table 4.2.1, 

‘‘Minimum Test and Calibration 

Frequency for Core Cooling, Rod Block 

and Isolation Instrumentation,’’ of the 

Technical Specifications to shorten the 

test interval between surveillance tests 

for the scram discharge volume high 

level rod block, and the safety/relief 

valve low-low set logic inhibit timer. 

Date of issuance: January 12, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment No.: 144. 

Facility Operating License No. DPR–

22. Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: January 18, 2005 (70 FR 

2892). The supplemental letters 

contained clarifying information and 

did not change the initial no significant 

hazards consideration determination 

and did not expand the scope of the 

original Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 12, 

2006. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of application for amendment: 

January 11, 2005. 

Brief description of amendment: The 

amendment deletes requirements from 

the Technical Specifications for annual 

Occupational Radiation Exposure 

Reports and Monthly Operating Reports. 

Date of issuance: January 11, 2006. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 161. 

Facility Operating License No. NPF–

57: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: March 29, 2005 (70 FR 

15946). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

January 11, 2006. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of application for amendment: 

February 25, 2005. 

Brief description of amendment: The 

amendment revised Technical 

Specification 3.1.3.1, ‘‘Control Rod 

Operability,’’ for the condition of having 

one or more scram discharge volume 

vents or drain lines with inoperable 

valves. 

Date of issuance: January 13, 2006. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 162. 

Facility Operating License No. NPF–

57: This amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: June 7, 2005 (70 FR 33217). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated January 13, 2006. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of application for amendments: 

January 11, 2005. 

Brief description of amendments: The 

amendments deleted requirements from 

the Technical Specifications (TSs) for 

annual Occupational Radiation 

Exposure Reports and Monthly 

Operating Reports. 

Date of issuance: January 11, 2006. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment Nos.: 270 and 251. 

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: March 29, 2005 (70 FR 15946) 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated January 11, 2006. 

No significant hazards consideration 

comments received: No 

PSEG Nuclear, LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of application for amendments: 

February 15, 2005. 

Brief description of amendments: 

These amendments delete the total 

water and steam volume of the reactor 

coolant system from TS 5.4.2. 

Date of issuance: January 11, 2006. 

Effective date: As of the date of 

issuance and to be implemented within 

60 days. 

Amendment Nos.: 269 and 250. 

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: March 29, 2005 (70 FR 

15940). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

January 11, 2006. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

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

Rhea County, Tennessee 

Date of application for amendment: 

April 4, 2005, as supplemented by 

letters dated September 30 and 

November 8, 2005. 

Brief description of amendment: The 

amendment supports the steam 

generator replacement project by 

temporarily allowing one of the shield 

building dome penetrations to be 

opened up to five hours a day, six days 

a week while in Modes 1–4 during 

Cycle 7 operation until entering Mode 5 

at the start of the Cycle 7 refueling 

outage in fall 2006. 

Date of issuance: January 6, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 59. 

Facility Operating License No. NPF–

90: Amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 19, 2005 (70 FR 41446). 

The supplemental letters provided 

clarifying information that was within 

the scope of the initial notice and did 

not change the initial proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 6, 2006. 

No significant hazards consideration 

comments received: No. 


hsrobinson on PROD1PC70 with NOTICES
Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 5087 

Notice of Issuance of Amendments to 

Facility Operating Licenses and Final 

Determination of No Significant 

Hazards Consideration and Opportunity 

for a Hearing (Exigent Public 

Announcement or Emergency 

Circumstances) 

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 for the 

amendment 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. 

Because of exigent or emergency 

circumstances associated with the date 

the amendment was needed, there was 

not time for the Commission to publish, 

for public comment before issuance, its 

usual Notice of Consideration of 

Issuance of Amendment, Proposed No 

Significant Hazards Consideration 

Determination, and Opportunity for a 

Hearing. 

For exigent circumstances, the 

Commission has either issued a Federal 

Register notice providing opportunity 

for public comment or has used local 

media to provide notice to the public in 

the area surrounding a licensee’s facility 

of the licensee’s application and of the 

Commission’s proposed determination 

of no significant hazards consideration. 

The Commission has provided a 

reasonable opportunity for the public to 

comment, using its best efforts to make 

available to the public means of 

communication for the public to 

respond quickly, and in the case of 

telephone comments, the comments 

have been recorded or transcribed as 

appropriate and the licensee has been 

informed of the public comments. 

In circumstances where failure to act 

in a timely way would have resulted, for 

example, in derating or shutdown of a 

nuclear power plant or in prevention of 

either resumption of operation or of 

increase in power output up to the 

plant’s licensed power level, the 

Commission may not have had an 

opportunity to provide for public 

comment on its no significant hazards 

consideration determination. In such 

case, the license amendment has been 

issued without opportunity for 

comment. If there has been some time 

for public comment but less than 30 

days, the Commission may provide an 

opportunity for public comment. If 

comments have been requested, it is so 

stated. In either event, the State has 

been consulted by telephone whenever 

possible. 

Under its regulations, the Commission 

may issue and make an amendment 

immediately effective, notwithstanding 

the pendency before it of a request for 

a hearing from any person, in advance 

of the holding and completion of any 

required hearing, where it has 

determined that no significant hazards 

consideration is involved. 

The Commission has applied the 

standards of 10 CFR 50.92 and has made 

a final determination that the 

amendment involves no significant 

hazards consideration. The basis for this 

determination is contained in the 

documents related to this action. 

Accordingly, the amendments have 

been issued and made effective as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the application for 

amendment, (2) the amendment to 

Facility Operating License, 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 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

The Commission is also offering an 

opportunity for a hearing with respect to 

the issuance of the amendment. 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, 

and electronically on the Internet at the 

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

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

are problems in accessing the document, 

contact the PDR Reference staff at 1 

(800) 397–4209, (301) 415–4737, or by email to pdr@nrc.gov. If a request for a 

hearing or petition for leave to intervene 

is filed by the above date, 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 identify 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 


hsrobinson on PROD1PC70 with NOTICES
5088 Federal Register / Vol. 71, No. 20 / Tuesday, January 31, 2006 / Notices 

1To the extent that the applications contain 

attachments and supporting documents that are not 

publicly available because they are asserted to 

contain safeguards or proprietary information, 

petitioners desiring access to this information 

should contact the applicant or applicant’s counsel 

and discuss the need for a protective order. 

opinion which support the contention 

and on which the petitioner intends to 

rely in proving the contention at the 

hearing. The petitioner must also 

provide references to those specific 

sources and documents of which the 

petitioner is aware and on which the 

petitioner 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.1

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 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. 

Each contention shall be given a 

separate numeric or alpha designation 

within one of the following groups: 

1. Technical—primarily concerns/ 

issues relating to technical and/or 

health and safety matters discussed or 

referenced in the applications. 

2. Environmental—primarily 

concerns/issues relating to matters 

discussed or referenced in the 

environmental analysis for the 

applications. 

3. Miscellaneous—does not fall into 

one of the categories outlined above. 

As specified in 10 CFR 2.309, if two 

or more petitioners/requestors seek to 

co-sponsor a contention, the petitioners/ 

requestors shall jointly designate a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. If a petitioner/requestor 

seeks to adopt the contention of another 

sponsoring petitioner/requestor, the 

petitioner/requestor who seeks to adopt 

the contention must either agree that the 

sponsoring petitioner/requestor shall act 

as the representative with respect to that 

contention, or jointly designate with the 

sponsoring petitioner/requestor a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. 

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. Since the Commission has 

made a final determination that the 

amendment involves no significant 

hazards consideration, if a hearing is 

requested, it will not stay the 

effectiveness of the amendment. Any 

hearing held would take place while the 

amendment is in effect. 

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 or 

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). 

Entergy Operations, Inc., Docket No. 

50–313, Arkansas Nuclear One, Unit 1 

(ANO–1), Pope County, Arkansas 

Date of amendment request: January 

3, 2006, as supplemented by letters 

dated January 6 and 10, 2006. 

Description of amendment request: 

Entergy Operations, Inc. (Entergy) 

requests an emergency Technical 

Specification (TS) change to the Steam 

Generator Level—Low allowable value 

of Limiting Condition for Operation 

3.3.11, ‘‘Emergency Feedwater [EFW] 

Initiation and Control (EFIC) System 

Instrumentation.’’ Operation at 100 

percent power with the current 

allowable value involves an increased 

risk of spurious EFW initiation. 

Therefore, Entergy requests a revised TS 

allowable value of ≥ 9.34 inches and a 

limiting trip setpoint value of ≥ 10.42 

inches in order to achieve and maintain 

100 percent power operation. An 

actuation time delay of ≤ 10.4 seconds is 

also proposed to minimize the 

possibility of inadvertent actuations 

during anticipated transients such as 

main feedwater transients or main 

turbine trips. 

Date of issuance: January 13, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 7 days from the date of issuance. 

Amendment No.: 227. 

Renewed Facility Operating License 

No. DPR–51: Amendment revised the 

Technical Specification. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): No. The 

Commission’s related evaluation of the 

amendment, finding of emergency 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated January 13, 

2006. 

Attorney for licensee: Nicholas S. 

Reynolds, Esquire, Winston and Stawn, 

1700 K Street, NW., Washington, DC 

20006–3817. 

NRC Branch Chief: David Terao. 

Dated at Rockville, Maryland, this 20th day 

of January 2006. 

For the Nuclear Regulatory Commission. 

Catherine Haney, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 06–744 Filed 1–30–06; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Draft NUREG–1824, ‘‘Verification & 

Validation of Selected Fire Models for 

Nuclear Power Plant Applications,’’

Draft for Comment 

AGENCY: Nuclear Regulatory 

Commission (NRC). 

ACTION: Notice of availability of Draft 

NUREG–1824, ‘‘Verification & 

Validation of Selected Fire Models for 

Nuclear Power Plant Applications’’ and 

request for public comment. 

SUMMARY: The NRC is announcing the 

availability of Draft NUREG–1824, 

‘‘Verification & Validation of Selected 

Fire Models for Nuclear Power Plant 

Applications Volumes 1 through 7,’’ for 

public comment. 

DATES: Comments on this document 

should be submitted by March 31, 2006. 


hsrobinson on PROD1PC70 with NOTICES