66570 Federal Register / Vol. 65, No. 215 / Monday, November 6, 2000 / Notices

records to ensure that an adequate level

of protection is provided to NRCclassified information and material.

Submit, by January 5, 2001, 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–1F23, Rockville, MD

20852. OMB clearance requests are

available at the NRC worldwide web

site: http://www.nrc.gov/NRC/PUBLIC/

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, Brenda Jo. Shelton, U.S. Nuclear

Regulatory Commission, T–6 E6,

Washington, DC 20555–0001, by

telephone at 301–415–7233, or by

Internet electronic mail at

BJS1@NRC.GOV.

Dated at Rockville, Maryland, this 30th day

of October, 2000.

For the Nuclear Regulatory Commission.

Brenda Jo. Shelton,

NRC Clearance Officer, Office of the Chief

Information Officer.

[FR Doc. 00–28357 Filed 11–3–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–461]

In the Matter of Amergen Energy

Company, LLC (Clinton Power

Station); Exemption

I

AmerGen Energy Company, LLC

(AmerGen, the licensee) is the holder of

Facility Operating License No. NPF–62

which authorizes operation of the

Clinton Power Station (CPS). 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 the licensee’s CPS

site in DeWitt County, Illinois.

II

The U.S. Nuclear Regulatory

Commission (NRC) has established

requirements in Appendix G of Part 50

to Title 10, Code of Federal Regulations

(10 CFR Part 50, Appendix G), to protect

the integrity of the reactor coolant

pressure boundary in nuclear power

plants. This Appendix to Part 50

requires the pressure-temperature (P–T)

limits for an operating plant to be at

least as conservative as those that would

be generated if the methods of

Appendix G to Section XI of the

American Society of Mechanical

Engineers Boiler and Pressure Vessel

Code (Appendix G to the Code) were

applied. The methodology of Appendix

G to the Code postulates the existence

of a sharp surface flaw in the reactor

pressure vessel (RPV) that is normal to

the direction of the maximum applied

stress. For materials in the beltline and

upper and lower head regions of the

RPV, the maximum flaw size is

postulated to have a depth that is equal

to one-fourth of the thickness and a

length equal to 1.5 times the thickness.

For the case of evaluating RPV nozzles,

the surface flaw is postulated to

propagate parallel to the axis of the

nozzle’s corner radius. The basic

parameter in Appendix G to the Code

for calculating P–T limit curves is the

stress intensity factor, Kl, which is a

function of the stress state and flaw

configuration. The methodology

requires that licensees determine the

reference stress intensity (Kla) factors,

which vary as a function of temperature,

from the reactor coolant system (RCS)

operating temperatures, and from the

adjusted reference temperatures (ARTs)

for the limiting materials in the RPV.

Thus, the critical locations in the RPV

beltline and head regions are the 1⁄4-

thickness (1⁄4T) and 3⁄4-thickness (3⁄4T)

locations, which correspond to the

points of the crack tips if the flaws are

initiated and grown from the inside and

outside surfaces of the vessel,

respectively. Regulatory Guide (RG)

1.99, Revision 2, provides an acceptable

method of calculating ARTs for ferritic

RPV materials; the methods of RG 1.99,

Revision 2, include methods for

adjusting the ARTs of materials in the

beltline region of the RPV, where the

effects of neutron irradiation may

induce an increased level of

embrittlement in the materials.

The methodology of Appendix G

requires that P–T curves must satisfy a

safety factor of 2.0 on primary

membrane and bending stresses during

normal plant operations (including

heatups, cooldowns, and transient

operating conditions), and a safety

factor of 1.5 on primary membrane and

bending stresses when leak rate or

hydrostatic pressure tests are performed

on the RCS. Table 1 to 10 CFR Part 50,

Appendix G, provides the staff’s criteria

for meeting the P–T limit requirements

of Appendix G to the Code and 10 CFR

Part 50, Appendix G.

By letter dated August 25, 2000, as

supplemented September 21, October

14, and October 25, 2000, AmerGen

submitted a license amendment request

to update the P–T limit curves for CPS.

In the submittals, AmerGen also

requested NRC approval for exemptions

to use Code Cases N–588 and N–640 as

methods that would allow AmerGen to

deviate from complying with the

requirements in 10 CFR Part 50,

Appendix G, for generating the P–T

limit curves.

Code Case N–588

AmerGen has requested, pursuant to

10 CFR 50.60(b), an exemption to use

Code Case N–588 as the basis for

evaluating the axial and circumferential

welds in the CPS RPV. The current

methods of appendix G to the Code

mandate consideration of an axial flaw

in full penetration RPV welds, and thus,

for circumferential welds, dictate that

the flaw be oriented transverse to the

axis of the weld. Postulation of an axial

flaw in a circumferential weld is

unrealistic because the length of the

flaw would extend well beyond the

girth of the circumferential weld and

into the adjoining base metal material.

Industry experience with the repair of

weld indications found during

preservice inspection, and data taken

from destructive examination of actual

vessel welds, confirms that any

remaining flaws are small, laminar in

nature, and do not transverse the weld

bead orientation. Therefore, any

potential 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. For circumferential RPV

welds, the methods of the Code Case

therefore postulate the presence of a

flaw that is oriented in a direction

parallel to the axis of the weld (i.e., in

a circumferential orientation).

An analysis provided to the American

Society of Mechanical Engineers

(ASME) Code’s Working Group on

Operating Plant Criteria (WGOPC) (in

which Code Case N–588 was developed)

indicated that if an axial flaw is

postulated on a circumferential weld,

then based on the correction factors for

Federal Register / Vol. 65, No. 215 / Monday, November 6, 2000 / Notices 66571

1The margin of safety of 4.18 is arrived at by

dividing 0.926 by 0.443 and then multiplying by the

required safety factor of 2.

2The Code Case accomplishes this by reducing

the Mm factors for circumferential welds that are

used for calculations of the stress intensities

attributed to primary membrane stresses (Klm) and

primary bending stresses (Klb). As stated previously,

for RPVs with wall thicknesses in the range of 4.0–

12.0 inches, the Mm factor for circumferential welds

is 0.443. This is the normal wall thickness range for

GE designed boiling water reactors.

3The most limiting 1⁄4T material for the

generation of the CPS P–T limits is Circumferential

Weld AE (Material Heat 76492). According to the

AmerGen submittal of August 25, 2000, this weld

has a 1⁄4T RTNDT value at 32 EFPY of 55°F.

Application of Code Case N–588 will change the

basis for evaluating the vessel to the next most

limiting plate or vertical weld material, which

according to AmerGen is material heat 3P4955

(used to fabricate vertical welds BE, BF, and BG,

which according to AmerGen have a 1⁄4T RTNDT

value at 32 EFPY of 51°F).

membrane stress (Mm) given in the Code

Case for the inside diameter

circumferential (0.443) and axial (0.926)

flaw orientations, it is equivalent to

applying a safety factor of 4.18 on the

pressure loading under normal

operating conditions.1 Appendix G to

the Code only requires that a safety

factor of 2 be placed on the contribution

of the pressure load in the case of an

axially-oriented flaw in an axial weld,

shell plate, or forging. By postulating a

circumferentially-oriented flaw on a

circumferential weld and using the

appropriate correction factor, the margin

of 2 is maintained for the stress integrity

calculation for the circumferential weld.

Consequently, the staff determined that

the postulation of an axially-oriented

flaw on a circumferential RPV weld

adds a level of conservatism in the P–

T limits that goes beyond the margins of

safety required by 10 CFR Part 50,

Appendix G, and by Appendix of the

Code. For this reason, the methods of

the Code Case reduce the applied stress

intensities for primary membrane and

bending stresses in circumferential

flaws by a factor of approximately 2

(≈0.926/0.443).2 This is realistic since

the postulated circumferential flaw in

the vessel will propagate if a stress is

applied in a direction normal to the axis

of the flaw (i.e., by application of an

axially oriented stress that results in

Mode I crack propagation of the

circumferential flaw). Such tensile

stresses in the RPVs are typically about

half the magnitudes of the

corresponding membrane stresses.

Application of Code Case N–588 will

only matter if the Code Case is applied

for the case where a circumferential

weld is the most limiting material in the

beltline region of the boiling water

reactor (BWR) designed RPV. Since

application of the Code Case methods

allow licensees to reduce the stress

intensities attributed to the

circumferential weld, the net effect of

the Code Case would allow AmerGen to

use the next most limiting base metal or

axial weld material in the RPV as the

basis for evaluating the vessel and

generating the P–T limit curves, if the

circumferential weld (girth weld) is the

most limiting material in the beltline

region of the vessel. In this case, the

Code Case is relevant to the evaluation

of the CPS RPV, because the CPS RPV

is limited by Circumferential Weld AE

(Material Heat 76492).3

WGOPC has concluded that

application of Code Case N–588 to plant

P–T limits are still sufficient to ensure

the structural integrity of RPVs during

plant operations. The staff has

concurred with WGOPC’s determination

and has previously granted exemptions

to use Code Case N–588 for the Quad

Cities Nuclear Power Station (NRC letter

to Commonwealth Edison dated

February 4, 2000). In the staff’s letter of

February 4, 2000, the staff concluded

that the procedure in Appendix G to the

Code was developed for axially oriented

flaws and that such a procedure was

physically unrealistic and overly

conservative for postulating flaws of this

orientation in a circumferential weld.

The staff also concluded that relaxation

of the requirements of Appendix G to

the Code by application of Code Case

N–588 is acceptable and would

maintain, 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 for the Quad Cities

RPVs and reactor coolant pressure.

AmerGen’s proposal to use Code N–588

for generation of the CPS P–T limit

curves is predicated on the same

technical basis as was used for

generation of the Quad Cities P–T

limits. The staff therefore concludes that

Code Case N–588 is acceptable for

application to the CPS P–T limits.

Hence, the staff concurs that relaxation

of the ASME Section XI, Appendix G,

requirements by application of ASME

Code Case N–588 is acceptable for CPS

and would maintain, 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.

Code Case N–640

AmerGen has requested, pursuant to

10 CFR 50.60(b), an exemption to use

ASME Code Case N–640 (previously

designated as Code Case N–626) as the

basis for establishing the P–T limit

curves. Code Case N–640 permits

application of the lower bound static

initiation fracture toughness value

equation (Klc equation) as the basis for

establishing the curves in lieu of using

the lower bound crack arrest fracture

toughness value equation (i.e., the Kla

equation, which is based on conditions

needed to arrest a dynamically

propagating crack, and which is the

method invoked by Appendix G to

Section XI of the ASME Code). 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. The Klc equation

appropriately implements the use of the

static initiation fracture toughness

behavior to evaluate the controlled

heatup and cooldown process of a

reactor vessel. The staff has required use

of 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. Therefore,

the lower bound static fracture

toughness Klc equation provides an

acceptable method for calculating P–T

limits. In addition, P–T curves based on

the Klc equation will enhance overall

plant safety by opening the P–T

operating window with the greatest

safety benefit in the region of low

temperature operations.

Generating the RCS P–T limit curves

developed in accordance with

Appendix G to the Code, without the

relief provided by ASME Code Case N–

640, would unnecessarily require the

RPV to be maintained at a temperature

exceeding 212 °F during the pressure

test. Consequently, steam vapor hazards

would continue to be one of the safety

concerns for personnel conducting

inspections in primary containment.

Implementation of the proposed curves,

as allowed by ASME Code Case N–640,

provides an adequate margin of safety

and would eliminate steam vapor

hazards by allowing inspections in

primary containment to be conducted at

a lower coolant temperature. Thus,

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

underlying purpose of the regulation

will continue to be served.

WGOPC has concluded that

application of Code Case N–640 to plant

P–T limits are still sufficient to ensure

the structural integrity of RPVs during

plant operations. The staff has

concurred with ASME’s determination

and has previously granted exemptions

to use Code Case N–640 for the Quad

66572 Federal Register / Vol. 65, No. 215 / Monday, November 6, 2000 / Notices

Cities Nuclear Power Station (NRC letter

to Commonwealth Edison dated

February 4, 2000). In the letter of

February 4, 2000, the staff concluded

that application of Code Case N–640

would not significantly reduce the

safety margins required by 10 CFR part

50, Appendix G, and would eliminate

steam vapor hazards by allowing

inspections in the primary containment

to be conducted at a lower coolant

temperature. The staff also concluded

that relaxation of the requirements of

Appendix G to the Code by application

of Code Case N–640 is acceptable and

would maintain, 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 for the Quad Cities

RPVs and reactor coolant pressure

boundary. AmerGen’s proposal to use

Code N–640 for generation of the CPS

P–T limit curves is predicated on the

same technical basis as was used for

generation of the Quad Cities P–T

limits. The staff therefore concludes that

Code Case N–640 is acceptable for

application to the CPS P–T limits.

Hence, the staff concurs that relaxation

of the ASME Section XI, Appendix G,

requirements by application of ASME

Code Case N–640 is acceptable for CPS

and would maintain, 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.

III

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

the exemption would be required to

approve the use of Code Cases N–588

and N–640. The staff examined the

licensee’s rationale to support the

exemption requests and concurred that

the use of the code cases would 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 Code; and Regulatory Guide

1.99, Revision 2, the staff concludes that

application of the code cases 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 exemption

under the special circumstances of 10

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

that the methodology of Code Cases N–

588 and N–640 may be used to revise

the P–T limits for Clinton Power

Station.

IV

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 AmerGen Energy Company, LLC,

exemption from the requirements of 10

CFR Part 50, Section 50.60(a) and 10

CFR Part 50, Appendix G, for Clinton

Power Station.

Pursuant to 10 CFR 51.32, an

environmental assessment and finding

of no significant impact has been

prepared and published in the Federal

Register (65 FR 61204). 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 30th day

of October 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing, Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–28358 Filed 11–3–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Nominations of New Members of the

Advisory Committee on the Medical

Uses of Isotopes

AGENCY: Nuclear Regulatory

Commission.

ACTION: Call for Nominations.

SUMMARY: The U.S. Nuclear Regulatory

Commission (NRC) is re-advertising for

nominations for the position of health

care administrator on the Advisory

Committee on the Medical Uses of

Isotopes (ACMUI).

DATES: Nominations are due on or

before January 5, 2001.

ADDRESSES: Submit nominations to: The

Office of Human Resources, Attn: Ms.

Joyce Riner, Mail Stop T2D32, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555.

FOR FURTHER INFORMATION, CONTACT:

Betty Ann Torres, Office of Nuclear

Material Safety and Safeguards, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Telephone:

301–415–0191.

SUPPLEMENTARY INFORMATION: The

ACMUI advises the NRC on policy and

technical issues that arise in regulating

the medical use of byproduct material.

Responsibilities include providing

comments on changes in NRC rules,

regulations, and guidance documents

concerning medical use; evaluating

certain non-routine uses of byproduct

material for medical use; providing

technical assistance in licensing,

inspection, and enforcement cases; and

bringing key issues to the attention of

NRC for appropriate action.

ACMUI members possess the medical

and technical skills needed to address

evolving issues. Currently, the ACMUI

membership consists of the following:

(a) Nuclear medicine physician; (b)

nuclear cardiologist; (c) medical

physicist in nuclear medicine unsealed

byproduct material; (d) a therapy

physicist; (e) a radiation safety officer;

(f) a nuclear pharmacist; (g) two

radiation oncologists; (h) health care

administrator; (i) patients’ rights and

care advocate; (j) Food and Drug

Administration representative; and (k)

state representative.

The NRC is inviting nominations for

the position of health care administrator

on the ACMUI. The term of the

individual currently occupying the

health care administrator position ends

September 30, 2001.

Nominees must include four copies of

their resumes, describing their

educational and professional

qualifications, and provide their current

addresses and telephone numbers.

Committee members serve a 3-year

term, with possible reappointment to an

additional 3-year term.

Nominees must be U.S. citizens and

be able to devote approximately 80

hours per year to committee business.

Members will be compensated and

reimbursed for travel (including perdiem in lieu of subsistence) and

secretarial and correspondence

expenses unless the member is a fulltime Federal employee. Full-time

Federal employees are only reimbursed

for travel expenses. Nominees will

undergo a security background check

and will be required to complete