Federal Register / Vol. 66, No. 87 / Friday, May 4, 2001 / Notices 22619

under 5 U.S.C. 552b(c), (4) and (6) of the

Government in the Sunshine Act.

Dated: May 1, 2001.

Susanne Bolton,

Committee Management Officer.

[FR Doc. 01–11292 Filed 5–3–01; 8:45 am]

BILLING CODE 7555–01–M

NATIONAL SCIENCE FOUNDATION

Proposal Review; Notice of Meetings

In accordance with the Federal

Advisory Committee Act (Pub. L. 92–

463, as amended), the National Science

Foundation (NSF) announces its intent

to hold proposal review meetings

throughout the year. The purpose of

these meetings is to provide advice and

recommendations concerning proposals

submitted to the NSF for financial

support. The agenda for each of these

meetings is to review and evaluate

proposals as part of the selection

process for awards. The majority of

these meetings will take place at NSF,

4201 Wilson Blvd., Arlington, Virginia

22230.

All of these meetings will be closed to

the public. The proposals being

reviewed include information of a

proprietary or confidential nature,

including technical information;

financial data, such as salaries; and

personal information concerning

individuals associated with the

proposals. These matters are exempt

under 5 U.S.C. 552b(c), (4) and (6) of the

Government in the Sunshine Act. NSF

will continue to review the agenda and

merits of each meeting for overall

compliance of the Federal Advisory

Committee Act.

These closed proposal review

meetings will no longer be announced

on an individual basis in the Federal

Register. NSF intends to publish a

notice similar to this on a quarterly

basis. For an advance listing of the

closed proposal review meetings that

include the names of the proposal

review panel and the time, date, place,

and any information on changes,

corrections, or cancellations, please visit

the NSF web-site: www.nsf.gov/home/

pubinfo/advisory.htm. This information

may also be requested by telephoning

703/292–8182.

Dated: May 1, 2001.

Susanne Bolton,

Committee Management Officer.

[FR Doc. 01–11287 Filed 5–3–01; 8:45 am]

BILLING CODE 7555–01–M

NUCLEAR REGULATORY

COMMISSION

[50–301]

Nuclear Management Company, LLC

Duane Arnold Energy Center;

Exemption

1.0 Background

Nuclear Management Company, LLC

(NMC, the licensee) is the holder of

Facility Operating License No. DPR–49

which authorizes operation of the

Duane Arnold Energy Center (DAEC).

The license provides, among other

things, that the facility is subject to all

rules, regulations, and orders of the U.S.

Nuclear Regulatory Commission (the

Commission) now or hereafter in effect.

The facility consists of a boiling water

reactor located on NMC’s DAEC site,

which is located in Linn County, Iowa.

2.0 Purpose

Title 10 of the Code of Federal

Regulations (10 CFR) 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 that, ‘‘The

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 P–

T limits must meet the safety margin

requirements specified in the American

Society of Mechanical Engineers

(ASME) Boiler and Pressure Vessel Code

(Code), Section XI, Appendix G.

To address provisions of the proposed

amendments to the technical

specification (TS) P–T limits, the

licensee requested in its submittal dated

October 16, 2000, that the staff exempt

DAEC from application of specific

requirements of 10 CFR part 50,

§ 50.60(a) and 10 CFR part 50, Appendix

G, and substitute use of ASME Code

Case N–640. Code Case N–640 permits

the use of an alternate reference fracture

toughness (Klc fracture toughness curve

instead of Kla fracture toughness curve)

for reactor vessel materials in

determining the P–T limits. The

proposed action is in accordance with

the licensee’s application for exemption

contained in the October 16, 2000,

submittal, and is needed to support the

TS amendment request that is contained

in the same submittal. The proposed

amendment will revise the P–T limits

for heatup, cooldown, and inservice test

limitations for the reactor coolant

system (RCS) to 25 and 32 effective full

power years (EFPYs).

Code Case N–640

The licensee has proposed an

exemption to allow use of ASME Code

Case N–640 in conjunction with ASME

Section XI, 10 CFR 50.60(a) and 10 CFR

part 50, Appendix G, to determine that

the P–T limits meet the underlying

intent of the Nuclear Regulatory

Commission (NRC) regulations.

The proposed amendment to revise

the P–T limits for DAEC relies in part

on the requested exemption. These

revised P–T limits have been developed

using the Klc fracture toughness curve

shown in ASME Section XI, Appendix

A, Figure A–2200–1, in lieu of the Kla

fracture toughness curve of ASME

Section XI, Appendix G, Figure G–

2210–1, as the lower bound for fracture

toughness. The other margins involved

with the ASME Section XI, Appendix G

process of determining P–T limit curves

remain unchanged.

Use of the Klc curve in determining

the lower bound fracture toughness in

the development of P–T operating limits

curve is more technically correct than

the Kla curve. The Klc curve

appropriately implements the use of

static initiation fracture toughness

behavior to evaluate the controlled

heatup and cooldown process of a

reactor vessel. The licensee has

determined that the use of the initial

conservatism of the Kla curve when the

curve was codified in 1974 was

justified. 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 curve is well beyond

the margin of safety required to protect

the public health and safety from

potential RPV failure. In addition, P–T

curves based on the Klc curve will

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 Appendix G of

ASME Section XI, and the minimum

required pressure for the reactor coolant

pump seals adjusted for instrument

uncertainties.

Since the RCS P–T operating window

is defined by the P–T operating and test

limit curves developed in accordance

with the ASME Section XI, Appendix G

procedure, continued operation of

DAEC with these P–T curves without

the relief provided by ASME Code Case

N–640 may unnecessarily restrict the P–

22620 Federal Register / Vol. 66, No. 87 / Friday, May 4, 2001 / Notices

T operating window, especially at low

temperature conditions. The operating

window becomes more restrictive with

continued reactor vessel service.

Implementation of the proposed P–T

curves, as allowed by ASME Code Case

N–640, does not significantly reduce the

margin of safety. Thus, pursuant to 10

CFR 50.12(a)(2)(ii), the underlying

purpose of the regulation will continue

to be served.

In summary, the ASME Section XI,

Appendix G procedure was

conservatively developed based on the

level of knowledge existing in 1974

concerning RPV materials and the

estimated effects of operation. Since

1974, the level of knowledge about these

topics has been greatly expanded. The

NRC staff concurs that this increased

knowledge permits relaxation of the

ASME Section XI, Appendix G

requirements by application of ASME

Code Case N–640, while maintaining,

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

underlying purpose of the ASME Code

and the NRC regulations to ensure an

acceptable margin of safety.

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. The staff

accepts the licensee’s determination that

an exemption would be required to

approve the use of Code Case N–640.

The staff examined the licensee’s

rationale to support the exemption

request and concurred that the use of

the code case would also meet the

underlying intent of these regulations.

Based upon a consideration of the

conservatism that is explicitly

incorporated into the methodologies of

10 CFR part 50, Appendix G; Appendix

G of the ASME Code; and regulatory

guide (RG) 1.99, Revision 2, the staff

concluded that application of the code

case as described would provide an

adequate margin of safety against brittle

failure of the RPV. This is also

consistent with the determination that

the staff has reached for other licensees

under similar conditions based on the

same considerations. Therefore, the staff

concludes that requesting the exemption

under the special circumstances of 10

CFR 50.12(a)(2)(ii) is appropriate and

that the methodology of Code Case N–

640 may be used to revise the P–T limits

for the DAEC RCS.

4.0 Conclusion

Accordingly, the Commission has

determined that, pursuant to 10 CFR

50.12(a), the exemption is authorized by

law, will not endanger life or property

or common defense and security, and is,

otherwise, in the public interest.

Therefore, the Commission hereby

grants NMC an exemption from the

requirements of 10 CFR part 50,

§ 50.60(a) and 10 CFR part 50, Appendix

G, for the DAEC.

Pursuant to 10 CFR 51.32, an

environmental assessment and finding

of no significant impact has been

prepared and published in the Federal

Register (66 FR 20692). Accordingly,

based upon the environmental

assessment, the Commission has

determined that the granting of this

exemption will not result in any

significant effect on the quality of the

human environment.

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 27th day

of April, 2001.

For the Nuclear Regulatory Commission.

Cynthia A. Carpenter,

Acting Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–11277 Filed 5–3–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 40–8964]

Rio Algom Mining Corp.

AGENCY: Nuclear Regulatory

Commission.

ACTION: Final finding of no significant

impact; Notice of opportunity for

hearing.

SUMMARY: The U.S. Nuclear Regulatory

Commission (NRC) proposes to renew

NRC Source Material License SUA–1548

to authorize the licensee, Rio Algom

Mining Corporation (RAMC), to

continue commercial operations of its in

situ leach (ISL) Rio Algom Smith Ranch

Uranium Recovery Project in Converse

County, Wyoming. This license

currently authorizes RAMC to receive,

acquire, possess, and transfer uranium

at the Rio Algom Smith Ranch Project,

which is located approximately 17 miles

(27 Kilometers) Northeast of Glenrock,

Wyoming. An Environmental

Assessment (EA) was performed by the

NRC staff in support of its review of

RAMC renewal request, in accordance

with the requirements of 10 CFR part

51. The conclusion of the

Environmental Assessment is a Finding

of No Significant Impact (FONSI) for the

proposed licensing action.

FOR FURTHER INFORMATION CONTACT: Mr.

John H. Lusher, Fuel Cycle Licensing

Branch, Division of Fuel Cycle Safety

and Safeguards, Office of Nuclear

Material Safety and Safeguards, U.S.

Nuclear Regulatory Commission, Mail

Stop T8–A33, Washington, DC 20555.

Telephone 301/415–7694.

SUPPLEMENTARY INFORMATION:

Background

At the Rio Algom Smith Ranch

Facility, ISL uranium recovery method

involves: (1) The injection of native

groundwater, with added sodium

carbonate/bicarbonate and oxygen or

hydrogen peroxide, into uraniumbearing orebody through injection wells;

(2) the chemical mobilization of the

uranium through oxidation and then

complexation with the carbonate

species; and (3) the extraction of the

uranium-bearing solution from the

subsurface through a pattern of

pumping wells. The uranium is

separated from the leach solution by

conventional ion exchange methods in

the processing facility. The resulting

uranium-poor solution is recharged with

carbonate and oxygen and returned to

the leaching zone for additional

uranium recovery. This cycle continues

until the ore zone is depleted or

recovery of uranium is no longer

economically feasible.

The recovered uranium solution is

processed further by using ammonia or

hydrogen peroxide to precipitate the

uranium into a slurry. The resulting

slurry is thickened by gravity settling,

and then washed and de-watered in a

filter press to about 50 percent solids.

The filter press solids (cake) are then

dried in a natural gas heated oil vacuum

dryer, to produce uranium oxide, which

is commonly known as ‘‘yellowcake.’’

The dried yellowcake is packaged in 55-

gallon (208-liter) steel drums for storage

and eventual shipment to a fuel

processing facility.

RAMC conducts uranium recovery

operations within designated areas

(wellfield units) of the Rio Algom Smith

Ranch site. These wellfield units consist

of about 50 acres (20 hectars) in size. A

number of well patterns are installed in

each wellfield unit, with each pattern

typically including four injection wells

laid out in a roughly rectangular shape

and one centrally-located pumping

(production) well. Currently, RAMC is

conducting uranium recovery

operations in three wellfield units.