24626 Federal Register / Vol. 72, No. 85 / Thursday, May 3, 2007 / Notices 

request to OMB and solicitation of 

public comment. 

SUMMARY: The NRC is preparing a 

submittal to OMB for review of 

continued approval of information 

collections under the provisions of the 

Paperwork Reduction Act of 1995 (44 

U.S.C. Chapter 35). 

Information pertaining to the 

requirement to be submitted: 

1. The title of the information 

collection: NRC Survey of Public 

Response to Emergencies. 

2. Current OMB approval number: 

3150-XXXX (New Collection). 

3. How often the collection is 

required: This is a one-time collection. 

4. Who is required or asked to report: 

Members of the public that reside 

within 10 mile Emergency Planning 

Zones of nuclear power plant. 

5. The estimated number of annual 

respondents: This is a one-time 

collection of 800 completed surveys. 

6. The number of hours needed 

annually to complete the requirement or 

request: 

One-time event. 277 hours 

((completed surveys 800 x .333 hrs per 

response = 267 hrs) + (uncompleted 

surveys 120 x .083 hrs per response = 

10 hrs)). 

7. Abstract: As part of NRC’s effort to 

review and improve emergency 

response program areas, a telephone 

survey will be conducted to assess the 

satisfaction of the pubic with existing 

protective action strategies, the 

effectiveness in which these strategies 

are conveyed to the public, and the 

public response to the possibility of 

modifying protection action strategies. 

The survey will produce statistical 

descriptions of customer satisfaction 

and acceptance of emergency response 

planning and protective actions. The 

response to the surveys will be used by 

the NRC in the development of new or 

modified protective action strategies 

including the types of strategies 

implemented and the means for which 

the information on protective actions 

may be disseminated to the public. The 

response may also support quality 

improvement in the existing emergency 

planning information in other areas 

indirectly related to protective actions. 

Submit, by July 2, 2007, comments 

that address the following questions: 

1. Is the proposed collection of 

information necessary for the NRC to 

properly perform its functions? Does the 

information have practical utility? 

2. Is the burden estimate accurate? 

3. Is there a way to enhance the 

quality, utility, and clarity of the 

information to be collected? 

4. How can the burden of the 

information collection be minimized, 

including the use of automated 

collection techniques or other forms of 

information technology? 

A copy of the draft supporting 

statement may be viewed free of charge 

at the NRC Public Document Room, One 

White Flint North, 11555 Rockville 

Pike, Room O–1 F21, Rockville, MD 

20852. OMB clearance requests are 

available at the NRC worldwide Web 

site: http://www.nrc.gov/public-involve/ 

doc-comment/omb/index.html. The 

document will be available on the NRC 

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, Margaret A. Janney, U.S. 

Nuclear Regulatory Commission, T–5 

F52, Washington, DC 20555–0001, by 

telephone at 301–415–7245, or by 

Internet electronic mail to 

INFOCOLLECTS@NRC.GOV. 

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. E7–8438 Filed 5–2–07; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Agency Information Collection 

Activities: Proposed Collection; 

Comment Request 

AGENCY: U.S. Nuclear Regulatory 

Commission (NRC). 

ACTION: Notice of pending NRC action to 

submit an information collection 

request to OMB and solicitation of 

public comment. 

SUMMARY: The NRC is preparing a 

submittal to OMB for review of 

continued approval of information 

collections under the provisions of the 

Paperwork Reduction Act of 1995 (44 

U.S.C. Chapter 35). 

Information pertaining to the 

requirement to be submitted: 

1. The title of the information 

collection: 10 CFR Part 35, ‘‘Medical 

Use of Byproduct Material’’. 

2. Current OMB approval number: 

3150–0010. 

3. How often the collection is 

required: Reports of medical events, 

doses to an embryo/fetus or nursing 

child, or leaking sources are reportable 

on occurrence. A certifying entity 

desiring to be recognized by the NRC 

must submit a one-time request for 

recognition. 

4. Who is required or asked to report: 

Physicians and medical institutions 

holding an NRC license authorizing the 

administration of byproduct material or 

radiation therefrom to humans for 

medical use. 

5. The number of annual respondents: 

8,751. 

6. The number of hours needed 

annually to complete the requirement or 

request: 987,764 hours. 

7. Abstract: 10 CFR Part 35, ‘‘Medical 

Use of Byproduct Material,’’ contains 

NRC’s requirements and provisions for 

the medical use of byproduct material 

and for issuance of specific licenses 

authorizing the medical use of this 

material. These requirements and 

provisions provide for the radiation 

safety of workers, the general public, 

patients, and human research subjects. 

10 CFR Part 35 contains mandatory 

requirements that apply to NRC 

licensees authorized to administer 

byproduct material or radiation 

therefrom to humans for medical use. 

The information in the required reports 

and records is used by the NRC to 

ensure that public health and safety is 

protected, and that the possession and 

use of byproduct material is in 

compliance with the license and 

regulatory requirements. 

Submit, by July 2, 2007, comments 

that address the following questions: 

1. Is the proposed collection of 

information necessary for the NRC to 

properly perform its functions? Does the 

information have practical utility? 

2. Is the burden estimate accurate? 

3. Is there a way to enhance the 

quality, utility, and clarity of the 

information to be collected? 

4. How can the burden of the 

information collection be minimized, 

including the use of automated 

collection techniques or other forms of 

information technology? 

A copy of the draft supporting 

statement may be viewed free of charge 

at the NRC Public Document Room, One 

White Flint North, 11555 Rockville 

Pike, Room O–1 F21, Rockville, MD 

20852. OMB clearance requests are 

available at the NRC worldwide Web 

site: http://www.nrc.gov/public-involve/ 

doc-comment/omb/index.html. The 

document will be available on the NRC 

home page site for 60 days after the 

signature date of this notice. 

Comments and questions about the 

information collection requirement may 

be directed to the NRC Clearance 

Officer, Margaret A. Janney (T–5 F52), 

U.S. Nuclear Regulatory Commission, 

Washington, DC 20555–0001, by 

telephone at 301–415–7245, or by 

Internet electronic mail to 

INFOCOLLECTS@NRC.GOV. 

mmaher on DSK3CLS3C1PROD with $$_JOB

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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