Federal Register / Vol. 72, No. 85 / Thursday, May 3, 2007 / Notices 24627 

Dated at Rockville, Maryland, this 26th day 

of April, 2007. 

For the Nuclear Regulatory Commission. 

Margaret A. Janney, 

NRC Clearance Officer, Office of Information 

Services. 

[FR Doc. E7–8439 Filed 5–2–07; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket No. 50–354] 

PSEG Nuclear LLC; Notice of 

Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The U.S. Nuclear Regulatory 

Commission (NRC or the Commission) 

is considering issuance of an 

amendment to Facility Operating 

License No. NPF–57 issued to PSEG 

Nuclear (the licensee) for operation of 

the Hope Creek Generating Station 

(Hope Creek) located in Salem County, 

New Jersey. 

The proposed amendment would 

increase the authorized maximum 

power level from 3339 megawatts 

thermal (MWt) to 3840 MWt, an 

increase of approximately 15 percent. 

Before issuance of the proposed 

license amendment, the Commission 

will have made findings required by the 

Atomic Energy Act of 1954, as amended 

(the Act), and the Commission’s 

regulations. 

The Commission has made a 

proposed determination that the 

amendment request involves no 

significant hazards consideration. Under 

the Commission’s regulations in Title 10 

of the Code of Federal Regulations (10 

CFR), Section 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. 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 CPPU [Constant Pressure Power 

Uprate] analyses, which were performed 

at or above CPPU power levels, 

included a review and evaluation of the 

structures, systems, and components 

(SSCs) that could be affected by the 

proposed change. The proposed 

amendment does not change the design 

function or operation of the affected 

SSCs. 

Plant specific analyses were 

performed in the following areas: 

Reactor Core and Reactor Internals (e.g., 

steam dryer), Reactor Coolant System 

and associated systems, Containment, 

Emergency Core Cooling Systems, 

Control and Instrumentation Systems, 

Electrical Systems, Balance of Plant 

Systems, and Radwaste Systems. The 

results of the analyses, which included 

evaluating the increase in the likelihood 

of an SSC malfunction, concluded that 

the SSCs are capable of performing their 

design functions at CPPU conditions. 

Comprehensive evaluations were 

performed on the steam dryer and other 

reactor internals for both operational 

and structural performance. Predicted 

steam dryer peak and alternating stress 

ratios remain within allowable levels. 

The existing margins to steam dryer 

alternating stress limits and the steam 

dryer monitoring program during power 

ascension provide assurance that steam 

dryer integrity will be maintained. 

Vibration evaluations at CPPU 

conditions were performed on the 

Reactor Internal components and 

Reactor Coolant and associated system 

piping. These included the Main Steam, 

Feedwater and Reactor Recirculation 

systems piping and supports. The 

results of the vibration analyses 

demonstrate that operation at CPPU 

conditions will not result in any 

detrimental effects. System values will 

remain within allowable American 

Society of Mechanical Engineers Boiler 

and Pressure Vessel Code (ASME Code) 

limits. In addition, the ASME Code and 

regulatory guidelines require vibration 

test data be taken on high-energy piping 

during initial CPPU startup. The 

vibration start-up test program will 

validate the vibration analyses that were 

performed, demonstrating adequate 

performance of the SSCs. 

Engineered Safety Features (ESF) 

were evaluated at CPPU conditions 

using NRC-approved methods. The 

Emergency Core Cooling Systems 

(ECCS) were evaluated to ensure they 

are capable of performing their design 

function during loss-of-coolantaccidents (LOCA). Adequate net 

positive suction head is maintained 

without reliance on post-accident 

containment pressure. CPPU does not 

result in an increase or decrease in the 

available water sources, and does not 

result in any change in the maximum 

nominal reactor operating pressure. The 

CPPU evaluations demonstrate that the 

ECCS performance satisfy the 

requirements of 10 CFR 50.46 and 10 

CFR [Part] 50 Appendix K. 

Balance-of-plant (BOP) systems and 

equipment were also evaluated for 

CPPU operation. The resulting 

evaluations demonstrate adequate 

performance with limited modifications 

that were or will be made to BOP 

components. 

These analyses, which included 

evaluating the increased likelihood of 

an SSC malfunction, confirm acceptable 

performance of plant SSCs under CPPU 

conditions. On this basis, PSEG 

concludes that there is no significant 

change in the ability of the SSCs to 

preclude or mitigate the consequences 

of accidents. 

The probability (frequency of 

occurrence) of postulated Design Basis 

Accidents (DBA), and other Updated 

Final Safety Analysis Report (UFSAR) 

evaluated accidents, occurring is not 

affected by the increased power level, 

and Hope Creek continues to comply 

with the regulatory and design basis 

criteria established for plant equipment. 

The changes in consequences of 

hypothetical accidents, which are 

assumed to occur at 102% of the CPPU 

RTP [Rated Thermal Power], compared 

to those previously evaluated, are in all 

cases insignificant. The CPPU accident 

evaluations do not exceed any of the 

NRC-approved acceptance limits. The 

spectrum of hypothetical accidents and 

transients has been investigated, and is 

shown to meet the plant’s currently 

licensed regulatory criteria. 

Consequently, there is no significant 

increase in the probability or 

consequences of an accident previously 

evaluated. 

The impact of CPPU on the 

radiological consequences of postulated 

DBAs, operational transients and other 

UFSAR accidents was evaluated. The 

magnitude of the potential 

consequences is dependent upon the 

quantity of fission products released to 

the environment, the atmospheric 

dispersion factors and the dose 

exposure pathways. The atmospheric 

dispersion factors and the dose 

exposure pathways are not changed by 

CPPU operation. The only factor which 

could influence the magnitude of the 

consequences is the quantity of activity 

released to the environment. For CPPU, 

the Control Rod Drop Accident (CRDA), 

Loss-of-Coolant Accident (LOCA), Fuel 

Handling Accident (FHA), Main 

Steamline Break Accident (MSLBA) and 

instrument line break accident (ILBA) 

were reanalyzed. 

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24628 Federal Register / Vol. 72, No. 85 / Thursday, May 3, 2007 / Notices 

The DBA that has historically been 

limiting from a radiological criterion is 

the LOCA, for which USNRC Regulatory 

Guide 1.183, Appendix A guidance was 

applied. Adherence to the guidance in 

RG 1.183, and the use of the specific 

values/limits contained in the Technical 

Specifications with as-tested postaccident performance of the safety grade 

engineered safety functions (ESF), 

provide the assurance for sufficient 

safety margin, including a margin to 

account for analysis uncertainties. The 

CPPU LOCA evaluation results include 

the 2% power uncertainty factor from 

Regulatory Guide 1.49. 

The results of the CPPU radiological 

analyses remain below the allowable 

limits of 10 CFR 50.67 and Table 6 in 

Regulatory Guide 1.183; the CPPU 

impact is minimal and all radiological 

limits are met at CPPU conditions. 

Therefore, the proposed change does not 

involve a significant increase in the 

radiological consequences of an 

accident previously evaluated. 

While the proposed CPPU 

amendment is not being submitted as a 

risk-informed licensing action, it was 

evaluated from a risk perspective using 

the NRC guidelines established in 

Regulatory Guide 1.174. Level 1 and 

Level 2 Probabilistic Risk Assessments 

(PRAs) were performed for the CPPU. 

When compared to the risk-acceptance 

guidelines presented in Regulatory 

Guide 1.174, the calculated changes in 

core damage frequency (CDF) and large 

early release frequency (LERF) are 

insignificant. Based on these results, 

PSEG concludes that the proposed 

amendment would not involve a 

significant increase in the probability of 

an accident previously evaluated. 

The impact of CPPU operation on 

plant operator actions and procedures 

was also evaluated. The operator action 

response times credited in the safety 

analyses in the UFSAR are not changed 

by CPPU. In addition, there is no change 

in Emergency Operating Procedure 

(EOP) strategy for CPPU operation. 

Based on the above, PSEG concludes 

that the proposed amendment would 

not involve a significant increase in the 

probability or consequences of an 

accident previously evaluated. 

2. Does the proposed change create 

the possibility of a new or different kind 

of accident from any accident 

previously evaluated? 

Response: No. 

As discussed above, the evaluation of 

the proposed amendment included 

review of the SSCs that could be 

affected by the proposed change. The 

proposed amendment does not change 

the design function or operation of the 

affected SSCs. The proposed 

amendment does not introduce any new 

or different plant safety-related 

equipment, and only involves 

instrument set-point changes for CPPU 

conditions, and minimal modifications 

to plant BOP power generation 

equipment. The proposed amendment 

does not significantly impact the 

manner in which the plant is operated, 

and does not have any significant 

impact on the capability the SCCs 

involved to perform their design 

function. 

No new operating mode, safetyrelated equipment lineup, accident 

scenario or equipment failure mode was 

identified. The CPPU evaluations also 

addressed the impact to postulated 

accidents, accident radiological 

consequences and operator response. No 

significant impacts were identified. The 

full spectrum of accident considerations 

has been evaluated, and no new, 

different, or limiting kind of accident 

has been identified. CPPU uses 

developed technology, and applies it 

within the capabilities of existing plant 

equipment in accordance with presently 

existing regulatory criteria to include 

NRC approved codes, standards and 

methods. The CPPU analyses results 

confirm acceptable performance of plant 

SSCs under CPPU conditions. 

Consequently, there are no new credible 

failure mechanisms, malfunctions, or 

accident initiators that were not 

previously evaluated in the plant design 

and licensing bases. 

Based on the preceding, PSEG 

concludes that the proposed change 

would not introduce any new or 

different kind of accident, or failure 

mode, not previously analyzed. 

3. Does the proposed change involve 

a significant reduction in a margin of 

safety? 

Response: No. 

Safety margins are applied to plant 

parameters to account for various 

uncertainties and to avoid exceeding 

regulatory and licensing limits. The 

proposed change does not involve a 

significant reduction in any margin of 

safety. First, due to continuing 

improvements in the analytical 

techniques (computer codes and data) 

based on several decades of BWR safety 

technology, plant performance feedback, 

and improved fuel and core designs, a 

significant increase has resulted in the 

design and operating margins between 

calculated safety analysis results and 

the licensing limits. These available 

safety analyses differences, combined 

with the excess as-designed equipment, 

system and component capabilities, 

provide BWR plants the capability to 

achieve an increase in their thermal 

power ratings within the existing design 

and licensing basis. The proposed CPPU 

will reduce some of the existing design 

and operational margins. However, 

safety margins are considered to not be 

significantly reduced if: (1) Applicable 

regulatory requirements, codes and 

standards or their alternatives approved 

for use by the NRC, are met, and (2) if 

safety analysis acceptance criteria in the 

licensing basis are met, or if proposed 

revisions to the licensing basis provide 

sufficient margin to account for analysis 

and data uncertainty. This is the case for 

the proposed CPPU amendment. 

Safety margin is related to the ability 

of the fission product barriers to limit 

the level of radiation dose to the public. 

The impact of the proposed CPPU 

amendment on the: (1) Fuel cladding 

barrier, (2) reactor coolant pressure 

boundary (RCPB) barrier, and (3) 

containment fission product barrier is 

discussed below. 

To assure that fuel cladding damage 

limits are not exceeded, the impact of 

the proposed amendment on fuel system 

design, nuclear system design, thermal 

and hydraulic design, accident and 

transient analyses, and fuel design 

limits was evaluated. No new fuel 

design, or change in the specified fuel 

design limits, is required for CPPU. The 

current fuel and core design limits will 

continue to be met; both the Safety 

Limit Minimum Critical Power Ratio 

(SLMCPR) and other applicable 

Specified Acceptable Fuel Design Limits 

(SAFDLs) are still met. Analyses for 

each fuel reload will continue to meet 

the criteria accepted by the NRC. 

Continued compliance with the 

SLMCPR and other SAFDLs will be 

confirmed on a cycle specific basis 

consistent with the criteria accepted by 

the NRC as specified in NEDO–24011, 

‘‘General Electric Standard Application 

for Reactor Fuel, GESTAR II.’’ The ECCS 

evaluation for CPPU demonstrates the 

continued conformance to the 

acceptance criteria of 10 CFR 50.46, for 

peak cladding temperature (PCT) and 

the other 10 CFR 50.46 parameters. The 

increased PCT consequences for CPPU 

are insignificant and remain 

substantially below the regulatory 

criteria. Therefore, the ECCS safety 

margin and fuel cladding margin (PCT) 

are not significantly impacted by CPPU. 

Challenges to the Reactor Coolant 

Pressure Boundary were evaluated at 

CPPU conditions (pressure, 

temperature, flow, and radiation) and 

were found to meet their acceptance 

criteria for allowable stresses and 

overpressure margin. These evaluations 

included (1) overpressure protection, (2) 

structural integrity of the RCPB piping, 

components, and supports, and (3) 

structural integrity of the reactor vessel. 

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Federal Register / Vol. 72, No. 85 / Thursday, May 3, 2007 / Notices 24629 

For the most limiting pressurization 

event, the peak calculated pressure 

remains below the ASME Code 

allowable peak pressure. The structural 

integrity of the RCPB piping, 

components, and supports was 

evaluated using NRC-approved 

methodology. The changes in flow, 

pressure and temperature associated 

with CPPU do not result in load limits 

being exceeded. Sufficient margin 

remains between the calculated stresses 

and ASME Code limits. In addition, the 

ASME Code and regulatory guidelines 

require vibration test data be taken on 

high-energy piping during initial CPPU 

startup. The vibration start-up test 

program will validate the vibration 

analyses that were performed, 

demonstrating adequate performance. 

The structural integrity of the reactor 

vessel was evaluated. The neutron 

fluence was re-analyzed in accordance 

with the requirements of 10 CFR [Part] 

50 Appendix G. The existing PressureTemperature (P–T) limit curves have 

been revised for CPPU conditions (a 

previous amendment to the Hope Creek 

license changed the P–T curves and 

included CPPU conditions). The reactor 

vessel materials surveillance program is 

unchanged by CPPU. The maximum 

normal operating reactor dome pressure 

for CPPU is unchanged and the vessel 

remains in compliance with regulatory 

requirements. Consequently, CPPU 

operation does not have an adverse 

effect on the reactor vessel fracture 

toughness. The structural evaluation of 

the vessel demonstrates that ASME 

Code requirements are met for normal, 

upset, emergency and accident 

conditions. 

Based on the preceding, PSEG 

concludes that the RCPB structural 

integrity will be maintained and the 

licensing basis requirements will 

continue to be met following 

implementation of the proposed CPPU. 

The impact of the proposed CPPU on 

the Containment was evaluated. The 

effect of CPPU on the peak values for 

containment pressure and temperature 

confirms the suitability of the plant for 

operation at CPPU RTP. Also, the effects 

of CPPU on the conditions that affect 

the containment dynamic loads were 

determined to be satisfactory for CPPU 

operation. Where plant conditions with 

CPPU are within the range of conditions 

used to define the current dynamic 

loads, current safety criteria are met and 

no further structural analysis was 

required. The change in short-term 

containment response is negligible. 

Because there will be more residual heat 

with CPPU, the containment long-term 

response slightly increases. However, 

containment pressures and temperatures 

remain below their design limits 

following any design basis accident, and 

thus, the containment and its cooling 

systems are satisfactory for CPPU 

operation. The small increase in the 

calculated post LOCA suppression pool 

temperature above the currently 

assumed peak temperature was 

evaluated and determined to be 

acceptable. Based on the use of 

conservative assumptions in these 

evaluations, PSEG concludes that 

containment structural integrity will be 

maintained under the proposed CPPU 

conditions, and the containment 

parameters will remain below design 

limits. Therefore there is no significant 

reduction in safety margin. 

In summary, challenges to the fuel, 

RCPB, and containment were evaluated 

for CPPU conditions. The structural 

integrity of the fission product barriers 

will be maintained under CPPU 

conditions. As such, the proposed 

amendment would not degrade 

confidence in the ability of the barriers 

to limit the level of radiation dose to the 

public. Fuel integrity is maintained by 

meeting existing design and regulatory 

limits. The calculated loads on all 

affected structures, systems and 

components, including the reactor 

coolant pressure boundary, will remain 

within their design allowables for all 

design basis event categories. The 

containment parameters remain below 

design limits. No NRC acceptance 

criterion will be exceeded. Because the 

Hope Creek configuration and responses 

to transients and hypothetical accidents 

do not result in exceeding the presently 

approved NRC acceptance limits, CPPU 

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. 

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. 

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 6D59, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

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

Documents may be examined, and/or 

copied for a fee, at the NRC’s Public 

Document Room (PDR), located at One 

White Flint North, Public File Area O1 

F21, 11555 Rockville Pike (first floor), 

Rockville, Maryland. 

The filing of requests for 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 O1F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

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24630 Federal Register / Vol. 72, No. 85 / Thursday, May 3, 2007 / Notices 

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 

requestors/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 

opinion which support the contention 

and on which the petitioner 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 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 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, 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. 

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(c)(1)(i)–(viii). 

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 Jeffrie J. Keenan, Esquire, 

Nuclear Business Unit—N21, P.O. Box 

236, Hancocks Bridge, NJ 08038, 

attorney for the licensee. 

For further details with respect to this 

action, see the application for 

amendment dated September 18, 2006, 

as supplemented by letters dated 

October 10, 2006, October 20, 2006, 

February 14, February 16, February 28, 

March 13, and April 18, 2007 which is 

available for public inspection at the 

Commission’s PDR, located at One 

White Flint North, File Public Area O1 

F21, 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. Persons who do not have 

access to ADAMS or who encounter 

problems in accessing the documents 

located in ADAMS, should contact the 

NRC PDR Reference staff by telephone 

at 1–800–397–4209, 301–415–4737, or 

by e-mail to pdr@nrc.gov. 

Dated at Rockville, Maryland, this 27th day 

of April 2007. 

For the Nuclear Regulatory Commission. 

James J. Shea, 

Project Manager, Plant Licensing Branch I– 

2, Division of Operating Reactor Licensing, 

Office of Nuclear Reactor Regulation. 

[FR Doc. E7–8437 Filed 5–2–07; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Sunshine Federal Register Notice 

DATE: Weeks of April 30, May 7, 14, 21, 

28, June 4, 2007. 

PLACE: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland. 

STATUS: Public and Closed. 

MATTERS TO BE CONSIDERED: 

Week of April 30, 2007 

There are no meetings scheduled for 

the Week of April 30, 2007. 

Week of May 7, 2007—Tentative 

Monday, May 7, 2007 

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

and State Materials and 

Environmental Management 

Programs (FSME) Programs, 

Performance, and Plans (Public 

Meeting) (Contact: George Deegan, 

301–415–7834). 

This meeting will be Web cast live at 

the Web address—http://www.nrc.gov. 

Week of May 14, 2007—Tentative 

There are no meetings scheduled for 

the Week of May 14, 2007. 

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