24410 Federal Register / Vol. 66, No. 93 / Monday, May 14, 2001 / Notices

calculated for a number of postulated

LOCAs of different sizes, locations, and

other properties sufficient to provide

assurance that the most severe

postulated LOCAs are calculated.

Appendix K of Part 50, ‘‘ECCS

Evaluation Models,’’ requires, among

other items, that the rate of energy

release, hydrogen generation, and

cladding oxidation from the metal/water

reaction shall be calculated using the

Baker-Just equation (Baker, L., Just, L.C.,

‘‘Studies of Metal Water Reactions at

High Temperatures, III. Experimental

and Theoretical Studies of the

Zirconium-Water Reaction,’’ ANL–6548,

page 7, May 1962) and implicitly

assumes that either zircaloy or ZIRLO

shall be used as the fuel rod cladding

material.

Sections 50.44, and 50.46, and

Appendix K of 10 CFR Part 50, make no

provisions for use of fuel rods clad with

other than zircaloy or ZIRLO. The

licensee has requested the use of

Framatome Cogema Fuels (FCF) ‘‘M5’’

advanced alloy for fuel rod cladding for

the TMI–1 Cycle 14 operation. In order

to accommodate the high fuel rod

burnups that are required for today’s

modern fuel management schemes and

core designs, FCF developed the M5

advanced fuel rod cladding and fuel

assembly structural material. The M5

alloy is a proprietary zirconium-based

alloy comprised of primarily zirconium

(∼99 percent) and niobium (∼1 percent).

The elimination of tin has resulted in

superior corrosion resistence and

reduced irradiation induced growth

relative to both standard Zircaloy and

low-tin Zircaloy. The addition of

niobium increases ductility. Since the

chemical composition of the M5 alloy

differs from the specifications of

Zircaloy or ZIRLO, a plant-specific

exemption is required to allow the use

of the M5 alloy as a fuel rod cladding

material at TMI–1. The M5 would also

be used for fuel assembly spacer grids,

fuel rod end plugs and fuel assembly

guide and instrument tubes.

3.0 Discussion

Pursuant to 10 CFR 50.12, the

Commission may, upon application by

any interested person or upon its own

initiative, grant exemptions from the

requirements of 10 CFR Part 50, when

(1) the exemptions are authorized by

law, will not present an undue risk to

public health or safety, and are

consistent with the common defense

and security; and (2) when special

circumstances are present. Special

circumstances are present whenever

application of the regulations in the

particular circumstances would not

serve the underlying purpose of the rule

or is not necessary to achieve the

underlying purpose of the rule.

The underlying purpose of Section

50.46 is to ensure that facilities have

adequate acceptance criteria for ECCS.

FCF demonstrated in its topical report,

BAW–10227P–A, ‘‘Evaluation of

Advanced Cladding and Structural

Material (M5) in PWR Reactor Fuel,’’

dated September 30, 1997, and

approved on February 4, 2000, that the

effectiveness of the ECCS will not be

affected by a change from zircaloy fuel

rod cladding to M5 fuel rod cladding.

The analysis described in BAW–

10227P–A also concludes that the ECCS

acceptance criteria applied to reactors

fueled with zircaloy fuel are also

applicable to reactors fueled with M5

fuel rod cladding.

The underlying purposes of Section

50.44 and Appendix K, paragraph I.A.5,

are to ensure that cladding oxidation

and hydrogen generation are

appropriately limited during a LOCA

and conservatively accounted for in the

ECCS evaluation model. Specifically,

Appendix K requires that the Baker-Just

equation be used in the ECCS evaluation

model to determine the rate of energy

release, cladding oxidation, and

hydrogen generation. In BAW–10227P–

A, FCF demonstrated that the Baker-Just

model is conservative in all post-LOCA

scenarios with respect to the use of M5

advanced alloy as fuel rod cladding

material. The licensee has stated that

the amount of hydrogen generated in an

M5-clad core will remain within the

TMI–1 design basis. The NRC staff has

reviewed the FCF’s advanced cladding

and structural material, M5, for

pressurized water reactor fuel

mechanical designs as described in

BAW–10227P–A. In its February 4,

2000, safety evaluation, the NRC staff

concluded that, to the extent specified

and with limitations noted in the NRC

staff’s evaluation, the assumptions

related to M5 material properties and

mechanical design methodology are

acceptable for referencing in fuel reload

licensing applications. The NRC staff

has determined that since the licensee

and FCF have ongoing processes which

assure that LOCA analysis input values

for peak cladding temperature-sensitive

parameters bound the as-operated plant

values for those parameters at TMI–1

and also have ongoing processes to

determine mixed-core penalties as

needed, the methodologies and analyses

described in BAW–10227P–A apply to

TMI–1 and the plant can be safely

operated within the bounds of those

analyses with mixed- and full-core

loadings of M5 clad fuel and other M5

core structures. The NRC staff further

concluded that since fuel assemblies

which utilize the two different alloys

(M5 and zircaloy) and which will be coresident in the core have only slight

geometry differences, there will be

virtually no thermal-hydraulic effect,

and a mixed core penalty in LOCA

evaluations would not have to be

assessed to compensate for the material

differences. Therefore, based on the

information described above, the NRC

staff has determined that the underlying

purposes of Section 50.44, and 50.46,

and Appendix K of 10 CFR Part 50 have

been achieved through the use of M5

advanced alloy as fuel rod cladding

material and core structure material,

and that the special circumstances

required by 10 CFR 50.12(a)(2)(ii) for

granting exemptions from 50.44, and

50.46, and Appendix K of 10 CFR Part

50 exist.

4.0 Conclusion

Accordingly, the Commission has

determined that, pursuant to 10 CFR

50.12(a)(ii), the exemption is authorized

by law, will not endanger life or

property or common defense and

security, and is, otherwise, in the public

interest. Also, special circumstances are

present. Therefore, the Commission

hereby grants AmerGen Energy

Company, LLC, an exemption from the

requirements of 10 CFR 50.44, 10 CFR

50.46, and Appendix K to 10 CFR Part

50, for Three Mile Island Nuclear

Station, Unit 1, as to the use of M5

cladding and core structures in lieu of

Zircaloy or ZIRLO as currently specified

or implied in those regulations.

Pursuant to 10 CFR 51.32, the

Commission has determined that the

granting of this exemption will not have

a significant effect on the quality of the

human environment (66 FR 23279).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 8th day

of May 2001.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–12035 Filed 5–11–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–272 and 50–311]

PSEG Nuclear LLC, Salem Nuclear

Generating Station, Unit Nos. 1 and 2;

Environmental Assessment and

Finding of No Significant Impact

The U.S. Nuclear Regulatory

Commission (NRC) is considering

Federal Register / Vol. 66, No. 93 / Monday, May 14, 2001 / Notices 24411

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 Nos. DPR–70 and DPR–75,

issued to PSEG Nuclear LLC (the

licensee), for operation of the Salem

Nuclear Generating Station (Salem),

Unit Nos. 1 and 2. The facility is located

at the licensee’s site on the southern end

of Artificial Island in Lower Alloways

Creek Township, Salem County, New

Jersey. Salem, New Jersey is located

approximately 7.5 miles northeast of the

site.

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

the licensee from implementing specific

requirements of 10 CFR Part 50,

Appendix G, for operation of Salem,

Unit Nos. 1 and 2. In conjunction with

the staff granting the proposed

exemption to the requirements of 10

CFR Part 50, Appendix G, the licensee

is proposing to substitute ASME Code

Case N–640, ‘‘Alternative Reference

Fracture Toughness for Development of

P/T Limit Curves for ASME Section XI,

Division I.’’

The proposed action is in accordance

with the licensee’s application for

exemption dated November 10, 2000, as

supplemented by letters dated March 28

and April 2, 2001.

The Need for the Proposed Action

ASME Code Case N–640 is needed to

revise the method used to determine the

reactor coolant system (RCS) P–T limits,

since continued use of the present

curves unnecessarily restricts the P–T

operating window. The methodology

currently used to determine the lower

bound fracture toughness of RPV

material for development of P–T limit

curves is based on the KIa fracture

toughness curve of ASME Section XI,

Appendix G, Figure G–2210–1. The

licensee has determined that the use of

the KIa curve provided appropriate

conservatism when it was codified in

1974 due to the limited knowledge of

RPV materials. However, since that

time, additional knowledge has been

gained about RPV materials, that

demonstrates that the lower bound on

fracture toughness provided by the KIa

curve is well beyond the margin of

safety required to protect the public

health and safety from potential RPV

failure. Implementation of ASME Code

Case N–640 would provide an

alternative to the methodology used to

develop P–T limit curves. The code case

methodology uses the KIc fracture

toughness curve shown in ASME

Section XI, Appendix A, Figure A–

2200–1, in lieu of the KIa fracture

toughness curve of ASME Section XI,

Appendix G. Other margins involved

with the ASME Section XI, Appendix G

process for establishing P–T limit curves

would remain unchanged. P–T curves

based on the KIc curve would enhance

overall plant safety by opening the P–T

operating window with the greatest

safety benefit in the region of low

temperature operations. The operating

window through which the operator

heats up and cools down the RCS is

determined by the difference between

the maximum allowable pressure

determined by ASME Section XI,

Appendix G, and the minimum required

pressure for the reactor coolant pump

(RCP) seals adjusted for instrument

uncertainties.

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 Case 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 alternative as described

is consistent with the intent of the

applicable regulations and would

provide an acceptable margin of safety

against brittle failure of the Salem, Unit

Nos. 1 and 2 RPV material.

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 impact.

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 Salem Nuclear

Generating Station, dated April 1973.

Agencies and Persons Consulted

In accordance with its stated policy,

on May 1, 2001, the staff consulted with

the New Jersey State official, Mr. R.

Pinney 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 November 10, 2000, as

supplemented by letters dated March 28

and April 2, 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

24412 Federal Register / Vol. 66, No. 93 / Monday, May 14, 2001 / Notices

the ADAMS Public Library component

on the NRC Web site, http:www.nrc.gov

(the Electronic Reading Room).

Dated at Rockville, Maryland, this 8th day

of May 2001.

For the Nuclear Regulatory Commission.

Robert J. Fretz,

Project Manager, Section 2, Project

Directorate I, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–12036 Filed 5–11–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Public Meeting on an Overview of

Activities Related to the Potential HighLevel Waste Repository at Yucca

Mountain, NV

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of public meeting in

Mesquite, Nevada.

SUMMARY: The U.S. Nuclear Regulatory

Commission (NRC) staff will hold a

public meeting on the high-level waste

repository licensing process. The

meeting is intended to foster a common

understanding among the stakeholders

on issues that would be associated with

the licensing process, should the U.S.

Department of Energy (DOE) submit a

license application to the NRC for a

possible geologic repository at Yucca

Mountain, Nevada. All meetings will be

facilitated by Francis X. Cameron,

Special Counsel for Public Liaison, of

the NRC Office of the General Counsel.

The meeting is primarily to acquaint

the public with the NRC’s high-level

waste licensing process. It will begin

with an overview of the licensing

process, followed by NRC presentations

on the role of the NRC technical staff in

evaluating the DOE license application,

and the NRC role with respect to the

transportation of high-level waste. An

opportunity for questions will be

provided. In addition, members of the

NRC staff will be available for informal

discussion with members of the public.

The time, date, and location of the

Public Meeting is shown below.

Time/Date: Thursday, May 24, 2001,

from 5:30 p.m.–7 p.m. (Pacific time).

Place: Mesquite City Hall, Council

Chamber, 10 E. Mesquite Boulevard,

Mesquite, Nevada.

FOR FURTHER INFORMATION CONTACT:

Francis X. Cameron, Special Counsel for

Public Liaison, Office of the General

Counsel, U.S. Nuclear Regulatory

Commission, Washington DC 20555–

0001, or by telephone: (301) 415–1642

or e-mail: fxc@nrc.gov.

Dated at Rockville, Maryland this 8th day

of May, 2001.

For the Nuclear Regulatory Commission,

Newton K. Stablein,

Acting Chief, High-Level Waste and

Performance Assessment Branch, Division of

Waste Management, Office of Nuclear

Material Safety and Safeguards.

[FR Doc. 01–12037 Filed 5–11–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Public Meetings on Issues Associated

With the Licensing Process for a

Possible High-Level Waste Repository

at Yucca Mountain, Nevada; Overview

of NRC’s Formal Hearing Process

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of public meetings in

Pahrump, Nevada and Las Vegas,

Nevada.

SUMMARY: In response to public

requests, the U.S. Nuclear Regulatory

Commission (NRC) staff will continue

its series of public meetings on the highlevel waste repository licensing process.

The next meetings are intended to foster

an understanding of the hearing process

that the NRC would use to decide

whether to issue a construction

authorization for a possible repository at

Yucca Mountain, Nevada, if the U.S.

Department of Energy (DOE) were to

submit a license application to the NRC.

On April 15, 2001, the Commission

announced its intent to retain a formal

hearing process for evaluating a

potential license application for a

geologic repository. Both meetings will

be facilitated by Francis X. Cameron,

Special Counsel for Public Liaison, of

the NRC Office of the General Counsel.

Two meetings on this topic will be

conducted to acquaint the public with

the NRC’s high-level waste hearing

process. They will begin with an

overview of the events that would have

to take place before NRC would initiate

a formal hearing, a general review of the

NRC’s licensing role, and a general

description of the NRC’s formal hearing

process. These presentations will be

followed by a question and answer

period. In addition, members of the NRC

staff will be available for informal

discussion with members of the public.

The time, date, and location of the

Public Meetings are shown below.

Time/Date: Tuesday, May 22, 2001,

from 6:30 p.m. to 9 p.m. (Pacific time).

Place: Mountain View Casino and

Bowl, 1750 Pahrump Valley Boulevard,

Pahrump, Nevada 89048.

Time/Date: Wednesday, May 23,

2001, from 1 p.m. to 4 p.m. (Pacific

time).

Place: Regional Transportation

Commission Building (Next to Clark

County Government Center) Room 108,

600 South Grand Central Parkway, Las

Vegas, Nevada 89155.

FOR FURTHER INFORMATION CONTACT:

Francis X. Cameron, Special Counsel for

Public Liaison, Office of the General

Counsel, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

001, or by telephone: (301) 415–1642 or

e-mail fxc@nrc.gov.

Dated at Rockville, Maryland this 8th day

of May 2001.

For the Nuclear Regulatory Commission.

C. William Reamer,

Chief, High-Level Waste Branch, Division of

Waste Management, Office of Nuclear

Material Safety and Safeguards.

[FR Doc. 01–12038 Filed 5–11–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Sunshine Act Meeting

AGENCY HOLDING THE MEETING: Nuclear

Regulatory Commission.

DATE: Weeks of May 14, 21, 28, June 4,

11, 18, 2001.

PLACE: Commissioner’s Conference

Room, 11555 Rockville Pike, Rockville,

Maryland.

STATUS: Public and Closed.

MATTERS TO BE CONSIDERED: 

Week of May 14, 2001

There are no meetings scheduled for the

Week of May 14, 2001.

Week of May 21, 2001—Tentative

There are no meetings scheduled for the

Week of May 21, 2001.

Week of May 28, 2001—Tentative

Wednesday, May 30, 2001

10:25 a.m. Affirmation Session (Public

Meeting) (If needed)

Week of June 4, 2001—Tentative

Tuesday, June 5, 2001

9:25 a.m. Affirmation Session (Public

Meeting) (If needed)

2 p.m. Discussion of Management Issues

(Closed-Ex. 2)

Wednesday, June 6, 2001

10:30 a.m. All Employees Meeting (Public

Meeting)

1:30 p.m. All Employees Meeting (Public

Meeting)