23278 Federal Register / Vol. 66, No. 89 / Tuesday, May 8, 2001 / Notices

Dated: May 3, 2001.

Beth M. McCormick,

Advisory Committee Management Officer,

National Aeronautics and Space

Administration.

[FR Doc. 01–11557 Filed 5–7–01; 8:45 am]

BILLING CODE 7510–01–M

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–338 and 50–339]

Virginia Electric and Power Company,

North Anna Power Station Units 1 and

2; Exemption

1.0 Background

The Virginia Electric and Power

Company (the licensee) is the holder of

Facility Operating Licenses NPF–4 and

NPF–7, which authorize operation of

the North Anna Power Station, Units 1

and 2. The licenses provide, among

other things, that the facility is subject

to all rules, regulations, and orders of

the U.S. Nuclear Regulatory

Commission (NRC, the Commission)

now or hereafter in effect.

The facility consists of two

pressurized-water reactors located in

Louisa County in the Commonwealth of

Virginia.

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 pressure or

leak testing conditions. Specifically, 10

CFR part 50, Appendix G states that

‘‘[t]he appropriate requirements on

* * * the pressure-temperature limits

and minimum permissible temperature

must be met for all conditions.’’

Appendix G of 10 CFR part 50 specifies

that the requirements for these limits are

the American Society of Mechanical

Engineers (ASME) Code, Section XI,

Appendix G Limits. RG 1.99, Rev. 2,

provides guidance for implementing 10

CFR Part 50, Appendix G. In GL 88–11,

the NRC staff advised licensees that the

staff would use RG 1.99, Rev. 2, to

review P–T limit curves. RG 1.99, Rev.

2, contains conservative methodologies

for determining the increase in

transition temperature and the decrease

in upper-shelf energy resulting from

neutron radiation.

To address provisions of amendments

to the technical specifications (TS)

regarding the P–T limits, low

temperature overpressure protection

(LTOP) system setpoints, and LTOP

system effective temperature (Tenable),

the licensee requested in its submittal

dated June 22, 2000, as supplemented

by letters dated September 19, 2000, and

January 4, February 14, March 13,

March 22, and April 11, 2001, that the

staff exempt North Anna Units 1 and 2

from application of specific

requirements of 10 CFR part 50,

appendix G, and substitute 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 KIa fracture

toughness curve) for reactor vessel

materials in determining the P–T limits,

LTOP system setpoints and Tenable, and

provides for plant-specific 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 plantspecific 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 and, therefore, an exemption to apply

the Code Case would be required by 10

CFR 50.60. 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 the licensee’s

analysis to use an edition and addenda

of Section XI of the ASME Code

incorporated by reference into 10 CFR

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).

The proposed amendments submitted

by the licensee will revise the P–T

limits of TS 3/4.4.9 related to the heatup

and cooldown of the reactor coolant

system (RCS), the LTOP system

setpoints and Tenable for the LTOP

system, for operation to 32.3 effective

full power years (EFPY) for Unit 1 and

34.3 EFPY for Unit 2.

ASME Code Case N–641

The licensee has proposed an

exemption to allow use of ASME Code

Case N–641 in conjunction with ASME

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

part 50, appendix G, to determine the P–

T limits, LTOP system setpoints and

Tenable.

The proposed amendments to revise

the P–T limits, LTOP system setpoints

and Tenable for North Anna Units 1 and

2 rely in part on the requested

exemption. The revised P–T limits,

LTOP system setpoints and Tenable have

been developed using the KIc fracture

toughness curve, in lieu of the KIa

fracture toughness curve, as the lower

bound for fracture toughness of the RPV

materials.

Use of the KIc curve in determining

the lower bound fracture toughness of

RPV steels is more technically correct

than use of the KIa curve 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 KIc curve appropriately

implements the use of static initiation

fracture toughness behavior to evaluate

the controlled heatup and cooldown

process of a reactor vessel. The staff has

required use of the conservatism of the

KIa curve since 1974, when the curve

was adopted by the ASME Code. This

conservatism was initially necessary

due to the limited knowledge of the

fracture toughness of RPV materials at

that time. Since 1974, additional

knowledge has been gained about RPV

materials, which demonstrates that the

lower bound on fracture toughness

provided by the KIa curve greatly

exceeds the margin of safety required to

protect the public health and safety

from potential RPV failure. In addition,

P–T curves, LTOP setpoints, and Tenable

based on the KIc 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.

Since an unnecessarily reduced P–T

operating window can reduce operator

flexibility without just basis,

implementation of the proposed P–T

curves, LTOP setpoints, and Tenable as

allowed by ASME Code Case N–641

may result in enhanced safety during

critical plant operational periods,

specifically heatup and cooldown

conditions. Thus, pursuant to 10 CFR

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

of 10 CFR 50.60 and appendix G to 10

CFR part 50 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

staff concurs that this increased

knowledge permits relaxation of the

ASME Section XI, Appendix G

requirements by application of ASME

Code Case N–641, while maintaining,

Federal Register / Vol. 66, No. 89 / Tuesday, May 8, 2001 / Notices 23279

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

underlying purpose of 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–641.

The staff examined the licensee’s

rationale to support the exemption

request and concurs that the use of the

Code case 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,

as discussed above, the staff concludes

that application of the Code case as

described would provide an adequate

margin of safety against brittle failure of

the RPV. This conclusion is also

consistent with the determinations that

the staff has reached for other licensees

under similar conditions based on the

same considerations.

Therefore, the staff concludes that

granting an exemption under the special

circumstances of 10 CFR 50.12(a)(2)(ii)

is appropriate and that the

methodologies of Code Case N–641 may

be used to revise the P–T limits, LTOP

setpoints, and Tenable for North Anna

Power Station, Units 1 and 2.

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 the licensee an exemption from

the requirements of 10 CFR part 50,

appendix G, for North Anna Power

Station, Units 1 and 2.

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 22018).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 2nd 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. 01–11567 Filed 5–7–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–289]

Amergen Energy Company, LLC; Three

Mile Island Nuclear Station, Unit 1

Environmental Assessment and

Finding of No Significant Impact

The U.S. Nuclear Regulatory

Commission (NRC) is considering

issuance of an exemption from certain

provisions of Sections 50.44 and 50.46

and Appendix K of Title 10 of the Code

of Federal Regulations (10 CFR) Part 50

for Facility Operating License No. DPR–

50, issued to AmerGen Energy

Company, LLC (the licensee), for

operation of the Three Mile Island

Nuclear Station, Unit 1 (TMI–1), located

in Dauphin County, Pennsylvania.

Environmental Assessment

Identification of the Proposed Action

The licensee requests an exemption

from the provisions of: (1) 10 CFR 50.44,

‘‘Standards for combustible gas control

system in light-water-cooled power

reactors,’’ which provide requirements

to control hydrogen generated by

Zircaloy or ZIRLO fuel cladding after a

postulated loss-of-coolant accident

(LOCA); (2) 10 CFR 50.46, ‘‘Acceptance

criteria for emergency core cooling

systems for light-water nuclear power

reactors,’’ which requires the calculated

emergency core cooling system (ECCS)

performance for reactors with Zircaloy

or ZIRLO fuel cladding to meet certain

criteria; and (3) Appendix K, ‘‘ECCS

Evaluation Models,’’ which presumes

the use of Zircaloy or ZIRLO fuel

cladding when doing calculations for

energy release, cladding oxidation, and

hydrogen generation after a postulated

LOCA.

The proposed action would allow the

licensee to generally use the M5

advanced alloy for fuel rod cladding in

fuel assemblies at TMI–1. Limited use of

M5 alloy in demonstration assemblies at

TMI–1 had previously been approved.

M5 alloy would also be used in fuel

assembly spacer grids and fuel rod end

plugs and fuel assembly guide and

instrument tubes. M5 alloy material

would be used in lieu of Zircaloy or

ZIRLO, the materials assumed to be

used in the cited regulations. The fuel

assemblies would be loaded into the

TMI–1 reactor core during the refueling

outage in the fall of 2001, and in use

during Cycle 14 and beyond operation.

The proposed action is in accordance

with the licensee’s application for

exemption dated December 20, 2000, as

supplemented by letter dated March 14,

2001.

The Need for the Proposed Action

Appendix K of 10 CFR part 50 and 10

CFR 50.46(a)(1)(i) require the

demonstration of adequate ECCS

performance for light-water reactors that

contain fuel consisting of uranium oxide

pellets enclosed in Zircaloy or ZIRLO

tubes. In addition, 10 CFR 50.44(a)

addresses requirements to control

hydrogen generated by Zircaloy or

ZIRLO fuel after a postulated LOCA.

Each of these three regulations, either

implicitly or explicitly assume that

either Zircaloy or ZIRLO is used as the

fuel rod cladding material. In order to

accommodate the high fuel rod burnups

that are required for modern fuel

management and core designs,

Framatome Cogema Fuels (FCF),

developed the M5 advanced fuel rod

cladding and fuel assembly structural

material. M5 is an alloy comprised

primarily of zirconium (∼99 percent)

and niobium (∼ 1 percent) that has

demonstrated superior corrosion

resistance and reduced irradiation

induced growth relative to both

standard and low-tin Zircaloy. However,

since the chemical composition of the

M5 advanced alloy differs from the

specifications of either Zircaloy or

ZIRLO, use of the M5 advanced alloy

falls outside of the strict interpretation

of these regulations. Therefore, approval

of this exemption request is needed to

permit the use of the M5 advanced alloy

as a fuel rod cladding material at TMI–

1. Limited use of the M5 alloy in

demonstration assemblies at TMI–1 had

previously been approved.

Pursuant to 10 CFR 50.12, the NRC

may grant exemptions which are

authorized by law, will not present an

undue risk to the health and safety of

the public, and are consistent with the

common defense and security, provided

that special circumstances are present.

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

Commission believes that special

circumstances are present whenever

application of the regulation in the

particular circumstances is not

necessary to achieve the underlying

purpose of the rule. The underlying

purpose of 10 CFR 50.46 is to ensure

that facilities have adequate acceptance

criteria for ECCS. FCF demonstrates in

its topical report BAW–10227P–A,

‘‘Evaluation of Advanced Cladding and