22018 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

Estimated Annual Burden Per

Voluntary Respondent: 30 hours.

Estimated Total Annual Burden on

Respondents: 150 hours.

Jacqueline Agtuca,

Chief of Staff.

[FR Doc. 01–10917 Filed 5–1–01; 8:45 am]

BILLING CODE 7565–01–P

NORTHEAST DAIRY COMPACT

COMMISSION

Notice of Meeting

AGENCY: Northeast Dairy Compact

Commission.

ACTION: Notice of meeting.

SUMMARY: The Compact Commission

will hold its regular monthly meeting to

consider matters relating to

administration and enforcement of the

price regulation. This meeting will be

held in Rhode Island, continuing the

Commission’s program of holding a

meeting in each of the Compact states.

In addition to receiving reports and

recommendations of its standing

Committees, the Commission will

receive a number of informational

reports, including reports on the

operation of the wholesale and retail

markets and about the impact of the

price regulation on the Rhode Island

WIC Program.

DATES: The meeting will begin at 10 a.m.

on Friday, May 11, 2001.

ADDRESSES: The meeting will be held at

the Newport Marriott Hotel, 25

America’s Cup Avenue, Newport, Rhode

Island.

FOR FURTHER INFORMATION CONTACT:

Daniel Smith, Executive Director,

Northeast Dairy Compact Commission,

64 Main Street, Room 21, Montpelier,

VT 05602. Telephone (802) 229–1941.

Authority: 7 U.S.C. 7256.

Dated: April 25, 2001.

Daniel Smith,

Executive Director.

[FR Doc. 01–10888 Filed 5–1–01; 8:45 am]

BILLING CODE 1650–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–338 and 50–339]

Virginia Electric and Power Company,

North Anna Power Station, Units 1 and

2; Environmental Assessment and

Finding of No Significant Impact

The U.S. Nuclear Regulatory

Commission (NRC) is considering

issuance of an exemption from the

requirements of Title 10 of the Code of

Federal Regulations (10 CFR) part 50,

appendix G, for Facility Operating

License Nos. NPF–4 and NPF–7, issued

to Virginia Electric and Power Company

(the licensee), for operation of the North

Anna Power Station, Units 1 and 2,

located in Louisa County, Virginia.

Environmental Assessment

Identification of the Proposed Action

10 CFR Part 50, Appendix G, requires

that the pressure-temperature (P–T)

limits be established for reactor pressure

vessels (RPVs) during normal operating

and hydrostatic or leak testing

conditions. Specifically, 10 CFR part 50,

Appendix G, states that ‘‘[t]he

appropriate requirements on both the

pressure-temperature limits and the

minimum permissible temperature must

be met for all conditions.’’ Appendix G

of 10 CFR part 50 specifies that the

requirements for these limits are

contained in the American Society of

Mechanical Engineers (ASME) Boiler

and Pressure Vessel Code (Code),

Section XI, Appendix G.

To address provisions of an

amendment to the Technical

Specifications P–T limits and lowtemperature overpressure protection

(LTOP) system setpoints, the licensee

requested in its submittal dated June 22,

2000, as supplemented on January 4,

February 14, March 13, and March 22,

2001, that the NRC staff exempt North

Anna Power Station from the

requirements of 10 CFR Part 50,

Appendix G, to allow the use of ASME

Code Case N–641.

Code Case N–641 permits the use of

an alternate reference fracture toughness

(KIC fracture toughness curve instead of

the KIA fracture toughness curve) for

reactor vessel materials in determining

the P–T limits, LTOP system setpoints

and Tenable, and provides for plantspecific evaluation of Tenable. Since the

KIC fracture toughness curve shown in

ASME Section XI, Appendix A, Figure

A–2200–1 (the KIC fracture toughness

curve) provides greater allowable

fracture toughness than the

corresponding KIa fracture toughness

curve of ASME Section XI, Appendix G,

Figure G–2210–1 (the KIa fracture

toughness curve), and a plant-specific

evaluation of Tenable would give lower

values of Tenable than use of a generic

bounding evaluation for Tenable, use of

Code Case N–641 for establishing the P–

T limits, LTOP system setpoints and

Tenable would be less conservative than

the methodology currently endorsed by

10 CFR Part 50, Appendix G. Although

the use of the KIC fracture toughness

curve in ASME Code Case N–641 was

recently incorporated into Appendix G

to Section XI of the ASME Code, an

exemption is still needed because 10

CFR Part 50, Appendix G requires a

licensee’s analysis to use an edition and

addenda of Section XI of the ASME

Code incorporated by reference into 10

CFR Part 50, section 50.55a, i.e., the

editions through 1995 and addenda

through the 1996 addenda (which do

not include the provisions of Code Case

N–641). Therefore, an exemption to

apply the Code case is required by 10

CFR Part 50, section 50.60. The

proposed action is in accordance with

the licensee’s application for exemption

dated June 22, 2000, as supplemented

by letters dated January 4, February 14,

March 13, and March 22, 2001.

The Need for the Proposed Action

ASME Code Case N–641 is needed to

revise the method used to determine the

reactor coolant system (RCS) P–T limits,

LTOP setpoints, and Tenable.

The purpose of 10 CFR part 50,

Section 50.60(a), and 10 CFR part 50,

appendix G, is to protect the integrity of

the reactor coolant pressure boundary in

nuclear power plants. This is

accomplished through these regulations

that, in part, specify fracture toughness

requirements for ferritic materials of the

reactor coolant pressure boundary.

Pursuant to 10 CFR part 50, appendix G,

it is required that P–T limits for the RCS

be at least as conservative as those

obtained by applying the methodology

of the ASME Code, Section XI,

Appendix G.

Current overpressure protection

system (OPPS) setpoints produce

operational constraints by limiting the

P–T range available to the operator to

heat up or cool down the plant. The

operating window through which the

operator heats up and cools down the

RCS becomes more restrictive with

continued reactor vessel service.

Reducing this operating window could

potentially have an adverse safety

impact by increasing the possibility of

inadvertent OPPS actuation due to

pressure surges associated with normal

plant evolutions such as reactor coolant

pump start and swapping operating

charging pumps with the RCS in a

water-solid condition. The impact on

the P–T limits and OPPS setpoints has

been evaluated for an increased service

period for operation to 32.3 effective

full-power years (EFPYs) for Unit 1 and

34.3 EFPYs for Unit 2, based on ASME

Code, Section XI, Appendix G

requirements. The results indicate that

these OPPS setpoints would

significantly restrict the ability to

perform plant heatup and cooldown,