Federal Register / Vol. 62, No. 151 / Wednesday, August 6, 1997 / Notices 42267

permit reasonable system makeup and

pressure control.

To prevent these difficulties, the

licensee has requested to use the ASME

Code Case N–514, ‘‘Low Temperature

Overpressure Protection,’’ which

designates the allowable pressure as 110

percent of that specified by 10 CFR part

50, Appendix G. This would provide an

increased band to permit system

makeup and pressure control. ASME

Code Case N–514 is consistent with

guidelines developed by the ASME

Working Group on Operating Plant

Criteria to define pressure limits during

LTOP events that avoid certain

unnecessary operational restrictions,

provide adequate margins against failure

of the reactor pressure vessel, and

reduce the potential for unnecessary

activation of pressure-relieving devices

used for LTOP. The content of this

ASME Code Case has been incorporated

into Appendix G of Section XI of the

ASME Code and published in the 1993

Addenda to Section XI and has been

incorporated into the latest draft of

Regulatory Guide 1.147 (Draft

Regulatory Guide DG–1050, Revision 12

of Regulatory Guide 1.147, Inservice

Inspection Code Case Applicability

ASME Section XI, dated May 1997).

However, 10 CFR 50.55a, ‘‘Codes and

Standards,’’ only authorizes addenda

through the 1988 Addenda.

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. According to 10 CFR

50.12(a)(2)(ii), special circumstances are

present whenever application of the

regulation in question is not necessary

to achieve the underlying purpose of the

rule.

The underlying purpose of 10 CFR

part 50, Appendix G, is to establish

fracture toughness requirements for

ferritic materials of pressure-retaining

components of the reactor coolant

pressure boundary to provide adequate

margins of safety during any condition

of normal operation, including

anticipated operational occurrences, to

which the pressure boundary may be

subjected over its service lifetime.

Section IV.A.2 of Appendix G requires

that the reactor vessel be operated with

P/T limits at least as conservative as

those obtained by following the

methods of analysis and the required

margins of safety of Appendix G of the

ASME Code.

Appendix G of the ASME Code

requires that the P/T limits be

calculated: (a) Using a safety factor of

two on the principal membrane

(pressure) stresses; (b) assuming a flaw

at the surface with a depth of onequarter (1⁄4) of the vessel wall thickness

and a length of six (6) times its depth;

and (c) using a conservative fracture

toughness curve that is based on the

lower bond of static, dynamic, and crack

arrest fracture toughness tests on

material similar to the Ginna reactor

vessel material.

In determining the setpoint for LTOP

events, the licensee proposed to use

safety margins based on an alternate

methodology consistent with the ASME

Code Case N–514 guidelines. The ASME

Code Case N–514 allows determination

of the setpoint for LTOP events such

that the maximum pressure in the vessel

would not exceed 110 percent of the P/

T limits of the existing ASME Code

Appendix G. 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

stress, the proposed criteria will provide

adequate margins of safety on 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.

IV

For the foregoing reasons, the NRC

staff has concluded that the licensee’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

application of 10 CFR 50.60 is not

necessary in order to achieve the

underlying purpose of this regulation.

Accordingly, the Commission has

determined that, pursuant to 10 CFR

50.12, this exemption is authorized by

law, will not present an undue risk to

the public health and safety, and is

consistent with the common defense

and security.

Accordingly, the Commission hereby

grants an exemption from 10 CFR 50.60

such that in determining the setpoint for

LTOP events, the Appendix G curves for

P/T limits are not exceeded by more

than 10 percent. This exemption

permits using the safety margins

recommended in the AMSE Code Case

N–514, in lieu of the safety margins

required by 10 CFR part 50, Appendix

G. This exemption is applicable only to

LTOP conditions during normal

operation.

Pursuant to 10 CFR 51.32, the

Commission has determined that the

granting of the exemption will have no

significant impact on the quality of the

human environment (62 FR 40554).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 28th day

of July, 1997.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 97–20643 Filed 8–5–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

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[Docket No. 50–223]

University of Massachusetts Lowell;

(University of Massachusetts Lowell

Research Reactor); Order Modifying

Facility Operating License No. R–125

I

The University of Massachusetts

Lowell (the licensee) is the holder of

Facility Operating License No. R–125

(the license) issued on December 24,

1974, by the U.S. Atomic Energy

Commission, and subsequently renewed

on November 21, 1985, by the U.S.

Nuclear Regulatory Commission (the

NRC or the Commission). The license

authorizes operation of the University of

Massachusetts Lowell Research Reactor

(the facility) at a power level up to 1

megawatt thermal (MW(t)). The facility

is a research reactor located in the

center of the North Campus of the

University of Massachusetts Lowell, in

the city of Lowell, Middlesex County, in

northeastern Massachusetts,

approximately 5 miles from the New

Hampshire border. The mailing address

is Radiation Laboratory, University of

Massachusetts Lowell, One University

Avenue, Lowell, Massachusetts 01854.

II

On February 25, 1986, the

Commission promulgated a final rule in

Section 50.64 of Title 10 of the Code of

Federal Regulations (10 CFR 50.64)

limiting the use of high-enriched

uranium (HEU) fuel in domestic
42268 Federal Register / Vol. 62, No. 151 / Wednesday, August 6, 1997 / Notices

research and test reactors (non-power

reactors) (see 51 FR 6514). The rule,

which became effective on March 27,

1986, requires that if Federal

Government funding for conversionrelated costs is available, each licensee

of a non-power reactor replace HEU fuel

at its facility with low-enriched

uranium (LEU) fuel acceptable to the

Commission unless the Commission has

determined that the reactor has a unique

purpose. The Commission issued the

rule to reduce the risk of theft and

diversion of HEU fuel used in nonpower reactors.

Paragraphs 50.64(b)(2)(i) and (ii)

require that a licensee of a non-power

reactor (1) not acquire more HEU fuel if

LEU fuel that is acceptable to the

Commission for that reactor is available

when the licensee proposes to acquire

HEU fuel and (2) replace all HEU fuel

in its possession with available LEU fuel

acceptable to the Commission for that

reactor in accordance with a schedule

determined pursuant to 10 CFR

50.64(c)(2).

Paragraph 50.64(c)(2)(i) requires,

among other things, that each licensee

of a non-power reactor authorized to

possess and to use HEU fuel develop

and submit to the Director of the Office

of Nuclear Reactor Regulation (Director)

by March 27, 1987, and at 12-month

intervals thereafter, a written proposal

for meeting the requirements of the rule.

The licensee shall include in its

proposal a certification that Federal

Government funding for conversion is

available through the U.S. Department

of Energy or other appropriate Federal

agency and a schedule for conversion,

based upon availability of replacement

fuel acceptable to the Commission for

that reactor and upon consideration of

other factors such as the availability of

shipping casks, implementation of

arrangements for available financial

support, and reactor usage.

Paragraph 50.64(c)(2)(iii) requires the

licensee to include in the proposal, to

the extent required to effect conversion,

all necessary changes to the license, to

the facility, and to licensee procedures.

This paragraph also requires the

licensee to submit supporting safety

analyses in time to meet the conversion

schedule.

Paragraph 50.64(c)(2)(iii) also requires

the Director to review the licensee

proposal, to confirm the status of

Federal Government funding, and to

determine a final schedule, if the

licensee has submitted a schedule for

conversion.

Section 50.64(c)(3) requires the

Director to review the supporting safety

analyses and to issue an appropriate

enforcement order directing both the

conversion and, to the extent consistent

with protection of public health and

safety, any necessary changes to the

license, the facility, and licensee

procedures. In the Federal Register

notice of the final rule (51 FR 6514), the

Commission explained that in most, if

not all, cases, the enforcement order

would be an order to modify the license

under 10 CFR 2.204.

Section 2.714 states the requirements

for a person whose interest may be

affected by any proceeding to initiate a

hearing or to participate as a party.

III

On May 21, 1993, as supplemented on

March 17, 1994, May 16, 1997, and June

6, 1997, the NRC staff received the

licensee’s conversion proposal,

including its proposed modifications

and supporting safety analyses. HEU

fuel elements are to be replaced with

LEU fuel elements. The fuel elements

contain fuel plates, typical of materials

test reactors, with the fuel meat

consisting of uranium silicide dispersed

in an aluminum matrix. These plates

contain the uranium-235 isotope at an

enrichment of less than 20 percent. The

NRC staff reviewed the licensee’s

proposal and the requirements of 10

CFR 50.64 and has determined that

public health and safety and common

defense and security require the

licensee to convert the facility from the

use of HEU to LEU fuel in accordance

with the attachment to this Order and

the schedule included herein. The

attachment to this Order specifies the

changes to the license conditions and

discusses the changes to Technical

Specifications that are needed to amend

the facility license.

IV

Accordingly, pursuant to Sections 51,

53, 57, 101, 104, 161b, 161i, and 161o

of the Atomic Energy Act of 1954, as

amended, and to Commission

regulations in 10 CFR 2.204 and 10 CFR

50.64, it is hereby ordered that:

Facility Operating License No. R–125

is modified by amending the license

conditions and technical specifications

as stated in the attachment to this Order

on the later date of either (1) the day the

licensee receives an adequate number

and type of LEU fuel elements to

operate the facility as specified in the

licensee proposal or (2) 30 days after the

date of publication of this Order in the

Federal Register.

V

Pursuant to the Atomic Energy Act of

1954, as amended, the licensee or any

other person adversely affected by this

Order may request a hearing within 30

days of the date of this Order. Any

request for a hearing shall be submitted

to the Director, Office of Nuclear

Reactor Regulation, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, with a copy to the

Assistant General Counsel for Hearings

and Enforcement at the same address. If

a person other than the licensee requests

a hearing, that person shall set forth

with particularity in accordance with 10

CFR 2.714 the manner in which his or

her interest is adversely affected by this

Order.

If a hearing is requested by the

licensee or a person whose interest is

adversely affected, the Commission

shall issue an order designating the time

and place of any hearing. If a hearing is

held, the issue to be considered at such

hearing is whether this Order should be

sustained.

This Order shall become effective on

the later date of either the day the

licensee receives an adequate number

and type of LEU fuel elements to

operate the facility as specified in the

licensee proposal or 30 days after the

date of publication of this Order in the

Federal Register or, if a hearing is

requested, on the date specified in an

order after further proceedings on this

Order.

Dated at Rockville, Maryland, this 31st day

of July 1997.

Attachment: As stated.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

Attachment to Order—Modifying Facility

Operating License No. R–125

A. License Conditions Revised and Added

by This Order.

2.B.(2) Pursuant to the Act and 10 CFR

part 70, ‘‘Domestic Licensing of Special

Nuclear Material,’’ to receive, possess, and

use at any one time up to 6.0 kilograms of

contained uranium-235 at enrichments equal

to or less than 20 percent in the form of

material test reactor (MTR) type reactor fuel

in connection with operation of the reactor

and 5 Ci Am–Be and 10 Ci Sb–Be neutron

sources for use in connection with operation

of the reactor.

2.B.(4) Pursuant to the Act and 10 CFR

part 70, ‘‘Domestic Licensing of Special

Nuclear Material,’’ to possess, but not use, up

to 4.80 kilograms of contained uranium-235

at greater than 20 percent enrichment in the

form of MTR-type reactor fuel until the

existing inventory of this fuel is removed

from the facility.

2.C.(2) Technical Specifications The

technical specifications contained in

Appendix A, as revised through Amendment

No. 12, are hereby incorporated in the

license. The licensee shall operate the facility

in accordance with the technical

specifications.
Federal Register / Vol. 62, No. 151 / Wednesday, August 6, 1997 / Notices 42269

2.C.(4) The licensee shall submit a startup

test report within six months of the initial

criticality with low-enriched uranium reactor

fuel in accordance with Amendment No. 12.

This report shall be sent as specified in 10

CFR 50.4, ‘‘Written Communications.’’

B. The technical specifications will be

revised by this Order in accordance with the

‘‘Enclosure to License Amendment No. 12,

Facility Operating License No. R–125, Docket

No. 50–223, Replacement Pages for Technical

Specifications,’’ and as discussed in the

safety evaluation for this Order.

[FR Doc. 97–20642 Filed 8–5–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Advisory Committee on Reactor

Safeguards Subcommittee Meeting on

Planning and Procedures; Notice of

Meeting

The ACRS Subcommittee on Planning

and Procedures will hold a meeting on

August 27, 1997, Room T–2B1, 11545

Rockville Pike, Rockville, Maryland.

The entire meeting will be open to

public attendance, with the exception of

a portion that may be closed pursuant

to 5 U.S.C. 552b(c) (2) and (6) to discuss

organizational and personnel matters

that relate solely to internal personnel

rules and practices of ACRS, and

information the release of which would

constitute a clearly unwarranted

invasion of personal privacy.

The agenda for the subject meeting

shall be as follows:

Wednesday, August 27, 1997—10:30

a.m. until 12:15 p.m.

The Subcommittee will discuss

proposed ACRS activities and related

matters. It may also discuss the

qualifications of candidates for

appointment to the ACRS. The purpose

of this meeting is to gather information,

analyze relevant issues and facts, and to

formulate proposed positions and

actions, as appropriate, for deliberation

by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff person named

below five days prior to the meeting, if

possible, so that appropriate

arrangements can be made.

Further information regarding topics

to be discussed, the scheduling of

sessions open to the public, whether the

meeting has been cancelled or

rescheduled, the Chairman’s ruling on

requests for the opportunity to present

oral statements, and the time allotted

therefor can be obtained by contacting

the cognizant ACRS staff person, Dr.

John T. Larkins (telephone: 301/415–

7360) between 7:30 a.m. and 4:15 p.m.

(EDT). Persons planning to attend this

meeting are urged to contact the above

named individual one or two working

days prior to the meeting to be advised

of any changes in schedule, etc., that

may have occurred.

Dated: July 31, 1997.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 97–20652 Filed 8–5–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Advisory Committee on Reactor

Safeguards Meeting of the ACRS

Subcommittee on Regulatory Policies

and Practices; Notice of Meeting

The ACRS Subcommittee on

Regulatory Policies and Practices will

hold a meeting on August 27, 1997,

Room T–2B3, 11545 Rockville Pike,

Rockville, Maryland.

The entire meeting will be open to

public attendance.

The agenda for the subject meeting

shall be as follows:

Wednesday, August 27, 1997—1:00 p.m.

until the conclusion of business

The Subcommittee will discuss

technical and policy issues regarding

the defense-in-depth approach to the

regulation of nuclear power plants and

its use in risk-informed regulation. The

purpose of this meeting is to gather

information, analyze relevant issues and

facts, and to formulate proposed

positions and actions, as appropriate,

for deliberation by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittee, along with

any of its consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the NRC staff,

and other interested persons regarding

this review.

Further information regarding topics

to be discussed, whether the meeting

has been cancelled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements,

and the time allotted therefor can be

obtained by contacting the cognizant

ACRS staff engineer, Dr. Medhat ElZeftawy (telephone 301/415–6889)

between 7:30 a.m. and 4:15 p.m. (EDT).

Persons planning to attend this meeting

are urged to contact the above named

individual one or two working days

prior to the meeting to be advised of any

potential changes to the agenda, etc.,

that may have occurred.

Dated: July 31, 1997.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 97–20653 Filed 8–5–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Advisory Committee on Reactor

Safeguards; Joint Meeting of the ACRS

Subcommittees on Probabilistic Risk

Assessment, Plant Operations, and on

Fire Protection; Notice of Meeting

The ACRS Subcommittees on

Probabilistic Risk Assessment, Plant

Operations, and on Fire Protection will

hold a joint meeting on August 28 and

29, 1997, Room T–2B3, 11545 Rockville

Pike, Rockville, Maryland.

The entire meeting will be open to

public attendance.

The agenda for the subject meeting

shall be as follows:

Thursday, August 28, 1997—8:30 a.m.

until the conclusion of business

Friday, August 29, 1997—8:30 a.m. until

the conclusion of business

The Subcommittees will continue

their review of matters included in the

Staff Requirements Memorandum (SRM)

dated May 27, 1997: (1) Acceptance

criteria for plant-specific safety goals

and deriving lower-tier acceptance

criteria; and (2) the use of uncertainty

versus point values in the PRA-related