Federal Register / Vol. 66, No. 122 / Monday, June 25, 2001 / Notices 33717

sites. With regard to thermal discharges

to the Susquehanna River, the staff has

previously evaluated temperature effects

during normal operations at full power

and determined the temperature impact

on the river to be insignificant. The

licensee indicated that an increase in

the cooling tower air flow rate will

compensate for the slight increase in

condenser outlet circulating water

temperature, such that no perceptible

change in the temperature of the cooling

tower basin blowdown to the

Susquehanna River is expected.

Therefore, the temperature effects on the

river will be insignificant. Existing

administrative controls ensure the

conduct of adequate monitoring such

that appropriate actions can be taken to

preclude exceeding the limits imposed

by the National Pollution Discharge

Elimination System permit. No

additional requirements or other

changes are required as a result of the

power uprate. No other non-radiological

impacts are associated with the

proposed action.

Based upon the above, the NRC

concludes that the proposed action does

not affect non-radiological plant

effluents and has no other

environmental impact. Therefore, there

are no significant non-radiological

environmental impacts associated with

the proposed action.

Accordingly, the NRC concludes that

there are no significant environmental

impacts associated with the proposed

action.

Alternatives to the Proposed Action

As an alternative to the proposed

action, the staff considered denial of the

proposed action (i.e., the ‘‘no-action’’

alternative). Denial of the application

would result in no change in current

environmental impacts. The

environmental impacts of the proposed

action and the alternative action are

similar.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement for the SSES, Units 1 and 2.

Agencies and Persons Consulted

In accordance with its stated policy,

on June 19, 2001, the staff consulted

with the Pennsylvania State official, Mr.

Michael Murphy of the Pennsylvania

Department of Environmental

Protection, regarding the environmental

impact of the proposed action. The State

official had no comments.

Finding of No Significant Impact

On the basis of the environmental

assessment, the NRC concludes that the

proposed action will not have a

significant effect on the quality of the

human environment. Accordingly, the

NRC has determined not to prepare an

environmental impact statement for the

proposed action.

For further details with respect to the

proposed action, see the licensee’s letter

dated October 30, 2000, as

supplemented by letters dated February

5, May 22, and May 31, 2001.

Documents may be examined, and/or

copied for a fee, at the NRC’s Public

Document Room, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland. Publicly

available records will be accessible

electronically 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/NRC/

ADAMS/index.html. If you do not have

access to ADAMS or if there are

problems in accessing the documents

located in ADAMS, contact the NRC

Public Document Room (PDR) Reference

staff at 1–800–397–4209, 301–415–4737

or by e-mail to pdr@nrc.gov.

Dated at Rockville, Maryland, this 19th day

of June 2001.

For the Nuclear Regulatory Commission.

Richard P. Correia,

Acting Chief, Section 1, Project Directorate

I, Division of Licensing Project Management,

Office of Nuclear Reactor Regulation.

[FR Doc. 01–15815 Filed 6–22–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–354]

PSEG Nuclear LLC; Hope Creek

Generating Station Environmental

Assesment and Finding of No

Significant Impact

The U.S. Nuclear Regulatory

Commission (NRC) is considering

issuance of an exemption from certain

requirements of Title 10 of the Code of

Federal Regulations (10 CFR) Part 50,

Appendix G, for Facility Operating

License No. NPF–57, issued to PSEG

Nuclear LLC, (the licensee) for operation

of the Hope Creek Generating Station

(HCGS), located in Salem County, New

Jersey.

Environmental Assessment

Identification of the Proposed Action

Title 10 of the Code of Federal

Regulations, Part 50, Appendix G,

requires that pressure-temperature (P–T)

limits be established for reactor pressure

vessels (RPVs) during normal operating

and hydrostatic or leak rate testing

conditions. Specifically, 10 CFR Part 50,

Appendix G, states, ‘‘The appropriate

requirements on both the pressuretemperature limits and the minimum

permissible temperature must be met for

all conditions.’’ The purpose of 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. Appendix G of 10 CFR Part

50 specifies that the requirements for

these limits are the American Society of

Mechanical Engineers (ASME) Boiler

and Pressure Vessel Code (Code),

Section XI, Appendix G Limits.

The proposed action would exempt

HCGS from application of specific

requirements of 10 CFR Part 50,

Appendix G, and would substitute use

of ASME Code Cases N–588 and N–640

as alternatives pursuant to 10 CFR

50.60(b).

The proposed action is in accordance

with the licensee’s application for

exemption dated December 1, 2000, as

supplemented by letters dated February

12, May 7, and May 14, 2001.

The Need for the Proposed Action

The proposed action is needed to

allow the licensee to implement ASME

Code Cases N–588 and N–640 in order

to revise the method used to determine

the P–T limits.

Code Case N–588, ‘‘Alternative to

Reference Flaw Orientation of Appendix

G for Circumferential Welds in Reactor

Vessels, Section XI, Division 1,’’ amends

the provisions of the 1989 Edition of

ASME Section XI, Appendix G, by

permitting the postulation of a

circumferentially oriented reference

flaw as the limiting flaw in a RPV

circumferential weld for the purpose of

establishing RPV P–T limits. The 1989

Edition of ASME Section XI, Appendix

G, would require that such a reference

flaw be postulated as an axially oriented

flaw in the circumferential weld. The

licensee addressed the technical

justification for this exemption by citing

industry experience and aspects of RPV

fabrication which support the

postulation of circumferentially

oriented flaws for these welds. The

reference flaw is a postulated flaw that

accounts for the possibility of a prior

existing defect that may have gone

undetected during the fabrication

process. Postulating the Appendix G

reference flaw in a circumferential weld

33718 Federal Register / Vol. 66, No. 122 / Monday, June 25, 2001 / Notices

is physically unrealistic and overly

conservative, because the length of the

flaw is 1.5 times the vessel wall, which

is much longer than the width of the

circumferential weld. Industry

experience with the repair of weld

indications found during preservice

inspection, inservice nondestructive

examinations, and data taken from

destructive examination of actual vessel

welds confirms that any remaining

defects are small, laminar in nature, and

do not cross transverse to the weld bead.

Therefore, any postulated defects

introduced during the fabrication

process, and not detected during

subsequent nondestructive

examinations, would only be expected

to be oriented in the direction of weld

fabrication. ASME Code Case N–588

also provides appropriate procedures for

determining the stress intensity factors

for use in developing RPV P–T limits

per ASME Code, Section XI, Appendix

G, procedures. The procedures allowed

by ASME Code Case N–588 are

conservative and provide a margin of

safety in the development of RPV P–T

operating and pressure test limits that

will prevent nonductile fracture of the

vessel.

Code Case N–640, ‘‘Alternative

Reference Fracture Toughness for

Development of P–T Limit Curves for

ASME Section XI, Division 1,’’ amends

the provisions of ASME Section XI,

Appendix G, by permitting the use of

the Klc equation as found in Appendix

A in ASME Section XI, in lieu of the Kla

equation as found in Appendix G in

ASME Section XI. Use of the Klc

equation in determining the lower

bound fracture toughness in the

development of the P–T operating limits

curve is more technically correct than

the use of the Kla equation since the rate

of loading during a heatup or cooldown

is slow and is more representative of a

static condition than a dynamic

condition. Use of Kla was justified by the

initial conservatism of the Kla equation

since 1974 when the equation was

codified. This initial conservatism was

necessary due to the limited knowledge

of RPV materials. Since 1974, additional

knowledge has been gained about RPV

materials, which demonstrates that the

lower bound on fracture toughness

provided by the Kla equation is well

beyond the margin of safety required to

protect the public health and safety

from potential RPV failure. The lower

bound Klc fracture toughness provides

an adequate margin of safety to protect

the public health and safety from

potential RPV failure.

The staff has determined that,

pursuant to 10 CFR 50.12(a)(2)(ii), the

underlying purpose of the regulation to

protect the integrity of the reactor

coolant pressure boundary will continue

to be served with the implementation of

Code Cases N–588 and N–640.

Environmental Impacts of the Proposed

Action

The NRC has completed its evaluation

of the proposed action and concludes

that the exemption and implementation

of the proposed alternatives as

described above are consistent with the

intent of the applicable regulations and

would provide an acceptable margin of

safety against brittle failure of the HCGS

RPV. Therefore, the proposed action

will not have a significant impact on the

environment.

The proposed action will not

significantly increase the probability or

consequences of accidents, no changes

are being made in the types of any

effluents that may be released offsite,

and there is no significant increase in

occupational or public radiation

exposure. Therefore, there are no

significant radiological environmental

impacts associated with the proposed

action.

With regard to potential

nonradiological environmental impacts,

the proposed action does not involve

any historic sites. It does not affect

nonradiological plant effluents and has

no other environmental impacts.

Therefore, there are no significant

nonradiological impacts associated with

the proposed action.

Accordingly, the NRC concludes that

there are no significant environmental

impacts associated with the proposed

action.

Alternatives to the Proposed Action

As an alternative to the proposed

action, the staff considered denial of the

proposed action (i.e., the ‘‘no-action’’

alternative). Denial of the application

would result in no change in current

environmental impacts. The

environmental impacts of the proposed

action and the alternative action are

similar.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement for the HCGS.

Agencies and Persons Consulted

In accordance with its stated policy,

on June 7, 2001, the staff consulted with

the New Jersey State official, Mr. Dennis

Zannoni, of the New Jersey Department

of Environmental Protection, regarding

the environmental impact of the

proposed action. The State official had

no comments.

Finding of No Significant Impact

On the basis of the environmental

assessment, the NRC concludes that the

proposed action will not have a

significant effect on the quality of the

human environment. Accordingly, the

NRC has determined not to prepare an

environmental impact statement for the

proposed action.

For further details with respect to the

proposed action, see the licensee’s letter

dated December 1, 2000, as

supplemented by letters dated February

12, May 7, and May 14, 2001.

Documents may be examined, and/or

copied for a fee, at the NRC’s Public

Document Room, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland. Publicly

available records will be accessible

electronically 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/NRC/

ADAMS/index.html. If you do not have

access to ADAMS or if there are

problems in accessing the documents

located in ADAMS, contact the NRC

Public Document Room (PDR) Reference

staff at 1–800–397–4209, 301–415–4737

or by e-mail to pdr@nrc.gov.

Dated at Rockville, Maryland, this 19th day

of June 2001.

For the Nuclear Regulatory Commission.

Richard B. Ennis,

Project Manager, Section 2, Project

Directorate I, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–15816 Filed 6–22–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Public Comment on Inspections, Tests,

Analyses and Acceptance Criteria

(ITAAC)

The U.S. Nuclear Regulatory

Commission (NRC) staff is seeking

public comment on ITAAC that are

required for issuance of combined

licenses (COLs) for nuclear power

facilities under Title 10 of the Code of

Federal Regulations (10 CFR) Part 52,

Subpart C. Subpart C of 10 CFR Part 52

sets forth a process for issuing combined

licenses (COLs) for nuclear power

facilities. A COL authorizes

construction and conditional operation

of a nuclear power facility. 10 CFR

Section 52.79(c) requires that the COL

application include ITAAC that are

necessary and sufficient to demonstrate

that the facility has been constructed

and will operate in conformity with the