3928 Federal Register / Vol. 63, No. 17 /Tuesday, January 27, 1998 / Notices

Addenda allows determination of the

setpoint for mitigating LTOP events so

that the maximum pressure in the vessel

would not exceed 110 percent of the

P–T limits that are determined using the

1996 methodology. This results in a

safety factor of 1.8 on the principal

membrane stresses. All other factors,

including assumed flaw size and

fracture toughness, remain the same.

Although this methodology would

reduce the safety factor on the principal

membrane stresses, the proposed

criteria will provide adequate margins

of safety for the reactor vessel during

LTOP transients and, thus, will satisfy

the underlying purpose of 10 CFR 50.60

for fracture toughness requirements.

Further, by relieving the operational

restrictions, the potential for

undesirable lifting of the PORV would

be reduced, thereby improving plant

safety.

It should be noted that the provision

to set the PORV setpoint so that system

pressure remains below 110 percent of

the P–T limits has already been

incorporated into the Byron and

Braidwood licensing basis. This

provision was approved by an

exemption to 10 CFR 50.60 granted to

Byron, Units 1 and 2, on November 29,

1996, to Braidwood, Unit 1 on July 13,

1995, and to Braidwood, Unit 2 on

December 12, 1997, to allow the use of

ASME Code Case N–514. Therefore,

although it represents a change from the

1989 Edition of the ASME Code, it is not

a change to the current licensing basis

for the facilities.

IV

For the foregoing reasons, the NRC

staff has concluded that ComEd’s

proposed use of the alternate

methodology in determining the

acceptable setpoint for LTOP events will

not present an undue risk to public

health and safety and is consistent with

the common defense and security. The

NRC staff has determined that there are

special circumstances present, as

specified in 10 CFR 50.12(a)(2), in that

10 CFR 50.60 need not be applied in

order to achieve the underlying purpose

of this regulation, which is to provide

adequate fracture toughness of the

reactor pressure boundary.

Accordingly, the Commission has

determined that, pursuant to 10 CFR

50.12(a), an 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 an exemption from the

requirements of 10 CFR 50.60 so that the

P–T limits may be determined using the

1996 Addenda to the ASME Code,

Section XI, Appendix G, and the LTOP

system setpoint may be determined so

that system pressure does not exceed

110 percent of the P–T limits.

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 (63 FR 2268).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 16th day

of January, 1998.

For the Nuclear Regulatory Commission.

Frank J. Miraglia,

Acting Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 98–1902 Filed 1–26–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–271]

Vermont Yankee Nuclear Power

Corporation, Vermont Yankee Nuclear

Power Station; Exemption

I

The Vermont Yankee Nuclear Power

Corporation (the licensee) is the holder

of Facility Operating License No. DPR–

28, which authorizes operation of the

Vermont Yankee Nuclear Power Station.

The license provides, among other

things, that the licensee is subject to all

rules, regulations, and orders of the

Nuclear Regulatory Commission (the

Commission) now or hereafter in effect.

The facility consists of a single-unit

boiling-water reactor located at the

licensee’s site in Windham County,

Vermont.

II

Section 70.24 of Title 10 of the Code

of Federal Regulations (10 CFR 70.24),

‘‘Criticality Accident Requirements,’’

requires that each licensee authorized to

possess special nuclear material (SNM)

shall maintain a criticality accident

monitoring system in each area where

such material is handled, used, or

stored. Subsections (a)(1) and (a)(2) of

10 CFR 70.24 specify detection and

sensitivity requirements that these

monitors must meet. Subsection (a)(1)

also specifies that all areas subject to

criticality accident monitoring must be

covered by two detectors. Subsection

(a)(3) of 10 CFR 70.24 requires licensees

to maintain emergency procedures for

each area in which this licensed SNM

is handled, used, or stored and also

requires that (1) the procedures ensure

that all personnel withdraw to an area

of safety upon the sounding of a

criticality accident monitor alarm, (2)

the procedures must include drills to

familiarize personnel with the

evacuation plan, and (3) the procedures

designate responsible individuals for

determining the cause of the alarm and

placement of radiation survey

instruments in accessible locations for

use in such an emergency. Subsection

(b)(1) of 10 CFR 70.24 requires licensees

to have a means for identifying quickly

personnel who have received a dose of

10 rads or more. Subsection (b)(2) of 10

CFR 70.24 requires licensees to

maintain personnel decontamination

facilities, to maintain arrangements for

the services of a physician and other

medical personnel qualified to handle

radiation emergencies, and to maintain

arrangements for the transportation of

contaminated individuals to treatment

facilities outside the site boundary.

Paragraph (c) of 10 CFR 70.24 exempts

Part 50 licensees from the requirements

of paragraph (b) of 10 CFR 70.24 for

SNM used or to be used in the reactor.

Paragraph (d) of 10 CFR 70.24 states that

any licensee who believes that there is

good cause why he or she should be

granted an exemption from all or part of

10 CFR 70.24 may apply to the

Commission for such an exemption and

shall specify the reasons for the relief

requested.

III

The SNM that could be assembled

into a critical mass at Vermont Yankee

is in the form of nuclear fuel; the

quantity of SNM other than fuel that is

stored on site in any given location is

small enough to preclude achieving a

critical mass. The Commission’s

technical staff has evaluated the

possibility of an inadvertent criticality

of the nuclear fuel at Vermont Yankee

and has determined that it is extremely

unlikely for such an accident to occur

if the licensee meets the following seven

criteria:

1. Only three new fuel assemblies are

allowed out of a shipping cask or

storage rack at one time.

2. The k-effective does not exceed

0.95, at a 95% probability, 95%

confidence level, in the event that the

fresh fuel storage racks are filled with

fuel of the maximum permissible U–235

enrichment and flooded with pure

water.

3. If optimum moderation occurs at

low moderator density, then the keffective does not exceed 0.98, at a 95%

probability, 95% confidence level, in

the event that the fresh fuel storage

racks are filled with fuel of the

maximum permissible U–235

enrichment and flooded with a
Federal Register / Vol. 63, No. 17 /Tuesday, January 27, 1998 / Notices 3929

moderator at the density corresponding

to optimum moderation.

4. The k-effective does not exceed

0.95, at a 95% probability, 95%

confidence level, in the event that the

spent fuel storage racks are filled with

fuel of the maximum permissible U–235

enrichment and flooded with pure

water.

5. The quantity of forms of SNM other

than nuclear fuel, that is stored on site

in any given area is less than the

quantity necessary for a critical mass.

6. Radiation monitors, as required by

General Design Criterion (GDC) 63, are

provided in fuel storage and handling

areas to detect excessive radiation levels

and to initiate appropriate safety

actions.

7. The maximum nominal U–235

enrichment is limited to 5.0 weight

percent.

By letter dated December 16, 1997,

the licensee requested an exemption

from 10 CFR 70.24. The licensee’s letter

dated January 13, 1998, provided

additional information supporting the

exemption. In the submittals, the

licensee addressed criteria 1, 2, 4, 5, 6,

and 7. Criterion 3 is satisfied because

the licensee’s submittal dated January

13, 1998, states that the cycle 20 fuel

will be channeled and stored in the

spent fuel storage pool until it is loaded

in the core and that the licensee has no

plans to store new fuel in the new fuel

storage vault. The Commission’s

technical staff has reviewed the

licensee’s submittals and has

determined that Vermont Yankee meets

the criteria for prevention of inadvertent

criticality; therefore, the staff has

determined that it is extremely unlikely

for an inadvertent criticality to occur in

SNM handling or storage areas at

Vermont Yankee.

The purpose of the criticality

monitors required by 10 CFR 70.24 is to

ensure that if a criticality were to occur

during the handling of SNM, personnel

would be alerted to that fact and would

take appropriate action. The staff has

determined that it is extremely unlikely

that such an accident could occur;

furthermore, the licensee has radiation

monitors that meet GDC 63 in fuel

storage and handling areas. These

monitors will alert personnel to

excessive radiation levels and allow

them to initiate appropriate safety

actions. The low probability of an

inadvertent criticality, together with the

licensee’s adherence to GDC 63,

constitutes good cause for granting an

exemption to the requirements of 10

CFR 70.24.

IV

The Commission has determined that

pursuant to 10 CFR 70.14, this

exemption is authorized by law, will not

endanger life or property or the common

defense and security, and is otherwise

in the public interest. Therefore, the

Commission hereby grants the Vermont

Yankee Nuclear Power Corporation an

exemption from the requirements of 10

CFR 70.24.

Pursuant to 10 CFR 51.32, the

Commission has determined that the

granting of this exemption will have no

significant impact on the human

environment (63 FR 2425).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 20th day

of January 1998.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 98–1901 Filed 1–26–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–22]

Westinghouse Electric Corporation

(CBS Corporation); Westinghouse Test

Reactor; Notice of Withdrawal of

Application for Consent to Transfer

Facility License and Conforming

Amendment

The U.S. Nuclear Regulatory

Commission (the Commission) has

permitted the withdrawal of the August

18, 1997 application for consent to

transfer Facility License No. TR–2 for

the Westinghouse Test Reactor, located

at the Westinghouse Waltz Mill site in

Westmoreland County, Pennsylvania,

and application for a conforming license

amendment; submitted by

Westinghouse Electric Corporation (CBS

Corporation).

The proposed action would have

approved the transfer of License No.

TR–2 from the Westinghouse Electric

Corporation to a new corporation that

would have taken the name

Westinghouse Electric Corporation, but

would not have included in its lines of

business certain media operations. The

proposed action would have also

amended the license to reflect the

proposed transfer of the license.

The Commission had previously

issued a Notice of Consideration of

Approval of Transfer of License and

Issuance of a Conforming Amendment

to Facility License, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for

Hearing published in the Federal

Register on September 26, 1997 (62 FR

50628). An Environmental Assessment

and Finding of No Significant Impact

was published in the Federal Register

on October 1, 1997 (62 FR 51493).

However, by letter dated December 18,

1997, the licensee withdrew the August

18, 1997 application.

The licensee withdrew the

application because its plan to

reorganize and create a new corporation

changed.

For further details with respect to this

action, see the application for

amendment dated August 18, 1997, and

the letter from licensee dated December

18, 1997, which withdrew the

application. The above documents are

available for public inspection at the

Commission’s Public Document Room,

the Gelman Building, 2120 L Street,

NW., Washington, DC.

Dated at Rockville, Maryland, this 20th day

of January 1998.

For the Nuclear Regulatory Commission.

Seymour H. Weiss,

Director, Non-Power Reactors and

Decommissioning Project Directorate,

Division of Reactor Program Management,

Office of Nuclear Reactor Regulation.

[FR Doc. 98–1899 Filed 1–26–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–263]

Draft Environmental Assessment;

Relating to a Proposed License

Amendment To Increase the Maximum

Rated Thermal Power Level at the

Monticello Nuclear Generating Plant

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of opportunity for public

comment.

SUMMARY: The Nuclear Regulatory

Commission has prepared a draft

environmental assessment related to the

Northern States Power Company’s

(NSP’s) request for a license amendment

to increase the maximum rated thermal

power level from 1670 megawattsthermal (MWt) to 1775 MWt. As stated

in the NRC staff’s position paper on the

Boiling-Water Reactor Extended Power

Uprate Program dated February 8, 1996,

the staff has the option of preparing an

environmental impact statement if it

believes a significant impact results

from the power uprate. The staff did not

identify a significant impact related to

the NSP’s request and, therefore, the

NRC staff documented its