39956 Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices

NW., Washington, DC, and accessible

electronically through the ADAMS

Public Electronic Reading Room link at

the NRC Web site (http://www.nrc.gov).

Dated at Rockville, Maryland this 23rd day

of June 2000.

For the Nuclear Regulatory Commission.

George F. Wunder,

Project Manager, Section 1, Project

Directorate 1, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–16296 Filed 6–27–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from June 3, 2000,

through June 16, 2000. The last

biweekly notice was published on June

14, 2000 (65 FR 37420).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to

4:15 p.m. Federal workdays. Copies of

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By July 28, 2000, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and electronically

from the ADAMS Public Library

component on the NRC Web site,

http://www.nrc.gov (the Electronic

Reading Room). If a request for a hearing

or petition for leave to intervene is filed

by the above date, the Commission or an

Atomic Safety and Licensing Board,

designated by the Commission or by the

Chairman of the Atomic Safety and

Licensing Board Panel, will rule on the

request and/or petition; and the

Secretary or the designated Atomic

Safety and Licensing Board will issue a

notice of a hearing or an appropriate

order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices 39957

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and electronically

from the ADAMS Public Library

component on the NRC Web site,

http://www.nrc.gov (the Electronic

Reading Room).

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of amendment request: June 14,

2000.

Description of amendment request:

The requested amendment proposes to

revise Technical Specification (TS) 5.6.5

to incorporate analytical methodologies

that are used for core operating limits

that have been accepted by NRC for

referencing in licensing applications.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Carolina Power & Light (CP&L) Company

has evaluated the proposed TS change and

has concluded that it does not involve a

significant hazards consideration. The

conclusion is in accordance with the criteria

set forth in 10 CFR 50.92. The bases for the

conclusion that the proposed change does

not involve a significant hazards

consideration are discussed below.

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed changes in a methodology

have been previously generically reviewed

and approved for use by the NRC for

determining core operating limits. Analyzed

events are assumed to be initiated by the

failure of plant structures, systems, or

components. The core operating limits

developed in accordance with the new

methodologies are bounded by the

limitations in the NRC acceptance in its

safety evaluations of the new methodologies.

The topical reports associated with the new

methodologies demonstrate that the integrity

of the fuel will be maintained during normal

operations and that design requirements will

continue to be met. The proposed change

does not have a detrimental impact on the

integrity of any plant structure, system, or

component. The proposed change will not

alter the operation of any plant equipment,

or otherwise increase its failure probability.

Therefore, the probability of occurrence for a

previously analyzed accident is not

significantly increased.

The consequences of a previously analyzed

accident are dependent on the initial

conditions assumed for the analysis, the

behavior of the fuel during the analyzed

accident, the availability and successful

functioning of the equipment assumed to

operate in response to the analyzed event,

and the setpoints at which these actions are

initiated. The proposed change to

methodology continues to meet applicable

design and safety analyses acceptance

criteria. The topical reports associated with

the new methodologies demonstrate that the

integrity of the fuel will be maintained as is

assumed or is bounded initially in accident

analyses. The proposed change does not

affect the performance of any equipment

used to mitigate the consequences of an

analyzed accident. As a result, no analyses

assumptions are violated and there are no

adverse effects on the factors that contribute

to offsite or onsite dose as the result of an

accident. The proposed change does not

affect setpoints that initiate protective or

mitigative actions. The proposed change

ensures that plant structures, systems, or

components are maintained consistent with

the safety analysis and licensing bases. Based

on this evaluation, there is no significant

increase in the consequences of a previously

analyzed event.

Therefore, the proposed change does not

involve any increase in the probability or

consequences of an accident previously

evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change does not involve any

physical alteration of plant systems,

structures, or components. The proposed

changes in methodology continue to meet

applicable criteria for MSLB [main steamline

break] and LBLOCA [large break loss-ofcoolant accident] analysis and assure that

appropriate criteria are used in future safety

analyses to establish the acceptability of

reload batch fuel with regard to mechanical

properties. The proposed change does not

involve a physical alteration of the plant

other than allowing for fuel design in

accordance with NRC approved

methodologies. No new or different

equipment is being installed. No installed

equipment is being operated in a different

manner. There is no alteration to the

parameters within which the plant is

normally operated or in the setpoints that

initiate protective or mitigative actions. As a

result no new failure modes are being

introduced. There are no changes in the

methods governing normal plant operation,

nor are the methods utilized to respond to

plant transients altered. Therefore, the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Does this change involve a significant

reduction in a margin of safety?

The margin of safety is established through

the design of the plant structures, systems,

and components, through the parameters

within which the plant is operated, through

the establishment of the setpoints for the

actuation of equipment relied upon to

respond to an event, and through margins

contained within the safety analyses. The

39958 Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices

proposed change in the methodologies used

for MSLB and LBLOCA analyses and the use

of the generic design criteria for PWR

[pressurized-water reactor] fuel designs does

not impact the condition or performance of

structures, systems, setpoints, and

components relied upon for accident

mitigation. The proposed change does not

significantly, impact any safety analysis

assumptions or results. Therefore, the

proposed change does not result in a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: William D.

Johnson, Vice President and Corporate

Secretary, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Section Chief: Richard P.

Correia.

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of amendment request: April 26,

2000.

Description of amendment request:

The proposed amendments would

revise Technical Specification Sections

3/4.3.7.1, ‘‘Radiation Monitoring

Instrumentation,’’ 3/4.7.2, ‘‘Control

Room and Auxiliary Electric Equipment

Room Emergency Filtration System,’’

and 6.2.F.8, ‘‘Ventilation Filter Testing

Program,’’ to eliminate habitability

system requirements associated with the

Auxiliary Electric Equipment Room

habitability systems.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Do the changes involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The elimination of Auxiliary Electric

Equipment Room (AEER) habitability system

requirements does not affect the precursors

or initiators of any accidents previously

evaluated.

The current analysis assumes an operator

will maintain continuous occupancy of the

AEER for 30 days following a design basis

loss-of-coolant-accident (LOCA). This

analysis credits operation of the AEER

habitability system. The resultant dose to the

operator is within the limits of 10 CFR 50,

Appendix A, ‘‘General Design Criteria for

Nuclear Power Plants,’’ General Design

Criterion (GDC) 19, ‘‘Control Room.’’ We

have performed an evaluation that

determined an operator has more than

sufficient time to perform all required actions

in the AEER following a design basis LOCA,

when directed by the station’s emergency

operating procedures (EOPs), without taking

credit for the AEER habitability system and

still maintain the resultant dose within the

limits of GDC 19.

Therefore, the proposed changes will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Do the changes create the possibility of a

new or different kind of accident from any

accident previously evaluated?

The proposed changes do not effect the

operation or configuration of plant systems,

structures, or components. These proposed

changes do not affect currently analyzed

failure modes and do not introduce new

failure modes.

Therefore, the proposed changes will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Do the changes involve a significant

reduction in a margin of safety?

The proposed changes will require an

operator to be present in the AEER in a postLOCA environment only when necessary to

perform required actions as directed by the

station’s EOPs. A time/motion study of

required AEER actions has determined that

the maximum cumulative time spent in the

AEER is approximately 300 minutes. The

dose to operators performing the required

AEER actions, without credit for the AEER

filtration system, will continue to be within

the limits of GDC 19, during and following

all design basis accidents.

Therefore, the proposed changes will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60690–0767.

NRC Section Chief: Anthony J.

Mendiola.

GPU Nuclear, Inc. and Saxton Nuclear

Experimental Corporation, Docket No.

50–146, Saxton Nuclear Experimental

Facility (SNEF), Bedford County,

Pennsylvania

Date of amendment request: April 10,

2000.

Description of amendment request:

The proposed amendment would make

changes to the organizational and

administrative controls for the SNEF to

reflect changes in GPU Nuclear, Inc.

following the sale of the Oyster Creek

Nuclear Generating Station. The

proposed changes to the technical

specifications (TSs) would (1) replace

reference to the President of GPU

Nuclear and division Vice Presidents

with a GPU Nuclear Cognizant Officer,

(2) replace reference to ‘‘other GPU

Nuclear personnel’’ with ‘‘other GPU

Inc, personnel,’’ (3) replace reference to

the ‘‘Radiation Safety Committee’’ with

the ‘‘TMI2/SNEC Oversight Committee,’’

(4) replace ‘‘GPU Nuclear audit program

procedures’’ with ‘‘approved Quality

Assurance Plan procedures,’’ and (5)

make changes to the TSs to reflect

changes to NRC organization.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

GPUN has determined that Technical

Specifications Change request No. 60

involves no significant hazards consideration

as defined in 10 CFR 50.92.

1. The proposed changes to the SNEC

Technical Specifications do not involve a

significant increase in the probability of

occurrence or consequences of an accident or

malfunction of equipment important to safety

previously analyzed in the safety analysis

report. The changes have no impact on plant

operations or the release of radioactive

materials.

2. The proposed changes to the SNEC

Technical Specifications will not create the

possibility for an accident or malfunction of

a different type than any previously

evaluated in the safety analysis report

because no plant configuration or operational

changes are involved.

3. The changes will not involve a

significant reduction in the margin of safety

as defined in the basis for any technical

specification for SNEC because no change to

operational limits will be made.

The NRC staff has reviewed the

analysis of the licensees and, based on

this review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for the Licensee: Ernest L.

Blake, Jr., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037.

NRC Branch Director: Ledyard B.

Marsh.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: February

3, 2000.

Description of amendment request:

NNECO’s proposed license amendment

Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices 39959

request of February 3, 2000, would add

a note to the Millstone 3 Final Safety

Analysis Report (FSAR) to indicate that

the configuration of relief valve

3CHS*V62 and isolation valve

3CHS*V61 takes exception to American

Society of Mechanical Engineers

(ASME) Section III code requirements

for class 2 components. The change

does not affect existing plant design but

rather changes licensing basis

information in the FSAR to accurately

reflect plant configuration.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff reviewed

the licensee’s analysis against the

standards of 10 CFR 50.92(c). The NRC

staff’s review is presented below:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The revision to the Final Safety Analysis

Report (FSAR) to correctly reflect the current

valve configuration to the Chemical Volume

and Control System (CVCS) will not affect

the ability of the CVCS to perform its

intended safety function. Therefore, this

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Since there are no changes in components,

component operation, or system operation,

this change does not create the possibility of

an accident of a different type.

3. Involve a significant reduction in a

margin of safety.

Since the FSAR revision does not have

anything to do with affecting the ability of

the CVCS to perform its intended safety

function, it will not involve a significant

reduction in a margin of safety.

Based on the staff’s analysis, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Section Chief: James W. Clifford.

Northern States Power Company,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of amendment request: May 12,

2000.

Description of amendment request:

The proposed amendment would revise

the Technical Specification Section

4.6.E.1.d safety/relief valve (SRV)

bellows monitoring system test

frequency from quarterly to once per

operating cycle.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment will have no

impact on the probability or consequences of

an accident. The BLDS [bellows leak

detection system] performs a monitoring

function only and is not part of the reactor

pressure boundary.

The reduced testing frequency for the leak

detection monitoring function will have no

impact on the ability of the pressure switch

to detect a bellows failure or on the

likelihood of bellows failure. Experience has

shown the pressure switch to be reliable and

capable of performing its function.

Reduction in test frequency to once per

cycle will still provide periodic verification

of pressure switch capability. Reduction in

test frequency to once per cycle will reduce

the number of times per cycle that SRV

operability is impacted by the testing

process. This will increase the probability

that SRV’s [sic] would be available to

mitigate consequences of an accident.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed amendment has the potential

to improve reliability of the BLDS by

removing a requirement which will allow

removal of a failure path. A reduction in

BLDS surveillance test frequency will not

result in creation of a new or different kind

of accident. The BLDS performs a monitoring

function only. It cannot cause an accident as

it is not part of the reactor pressure

boundary.

3. The proposed amendment will not

involve a significant reduction in the margin

of safety.

Revising the requirement to test this

system from quarterly to once per cycle will

not reduce the margin of safety. The pressure

switch and pressure boundary components of

the BLDS are reliable and stable. Therefore,

the proposed Technical Specification change

does not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jay E. Silberg,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, NW, Washington, DC

20037.

NRC Section Chief: Claudia M. Craig.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Units Nos.

1 and 2, Salem County, New Jersey

Date of amendment request: March 2,

2000.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3/4.6.3,

‘‘Containment Isolation Valves.’’ The

proposed change deletes the asterisk (*)

modifying the word OPERABLE in the

Limiting Condition for Operation and

relocates its associated footnote at the

bottom of the page to immediately

following the Action Statement. The

new note would be reworded to be

consistent with the wording of NUREG–

1431, ‘‘Standard Technical

Specifications, Westinghouse Plants.’’

The Bases associated with this TS

would also be revised to address the

proposed change.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The current Salem Technical

Specifications allows the use of

administrative means to unisolate a

containment isolation valve on an

intermittent basis. The proposed change

eliminates the potential for varying

interpretations of the TS footnote by

relocating it to the ACTION section of the

Technical Specifications in accordance with

the guidance of NUREG 1431, Rev 1 (April

1995) ‘‘Standard Technical Specifications

Westinghouse Plants (NUREG–1431).’’

PSE&G [PSE&G] views the proposed change

as a change that is editorial in nature.

The proposed change does not delete any

existing surveillance requirements or delete

any requirements from the Limiting

Condition for Operations (LCOs) or Action

Statements, and therefore does not reduce the

actions that are currently taken in the TS to

demonstrate operability of plant structures,

systems, or components (SSCs). The

proposed change continues to ensure the

operability of the containment isolation

valves, therefore ensuring that the

containment atmosphere will be isolated

from the outside environment in the event of

a release of radioactive material to the

containment atmosphere or pressurization of

the containment.

Since these changes do not modify any

SSCs or reduce the current requirements for

demonstrating operability of these SSCs, the

proposed changes to the TS do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated in the Safety Analysis Report

(SAR).

39960 Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed amendment eliminates the

potential for varying interpretations of the TS

footnote by relocating it to the ACTION

section of the Technical Specifications in

accordance with the guidance of NUREG

1431, Rev 1 (April 1995) ‘‘Standard

Technical Specifications Westinghouse

Plants (NUREG–1431).’’

The proposed change does not alter the

physical configuration of the plant. The

proposed change does not affect any systems,

structures or components assumed to

function in the accident analysis, or creates

a new or different accident scenario. The

proposed change to the TS does not affect the

ability of the plant systems to meet their

current TS requirements or design basis

functions. Therefore, the proposed change

does not increase the consequences of a

malfunction of equipment important to safety

previously evaluated in the SAR or create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed amendment eliminates the

potential for varying interpretations of the TS

footnote by relocating it to the ACTION

section of the Technical Specifications in

accordance with the guidance of NUREG

1431, Rev 1 (April 1995) ‘‘Standard

Technical Specifications Westinghouse

Plants.’’ The proposed amendment does not

change any testing acceptance criteria or

modify any protective trip setpoint. The

proposed change will continue to ensure that

the containment atmosphere will be isolated

from the outside environment in the event of

a release of radioactive material to the

containment atmosphere or pressurization of

the containment.

There is no reduction in the current

surveillance requirements required to

demonstrate the operability of plant SSCs.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jeffrie J. Keenan,

Nuclear Business Unit—N21, P.O. Box

236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Units Nos.

1 and 2, Salem County, New Jersey

Date of amendment request: April 13,

2000.

Description of amendment request:

The proposed amendments would

delete Technical Specification (TS) 3/

4.1.3.2.2 which is related to shutdown

and control rod group demand position

indication in modes 3, 4, and 5.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS change does not involve

any physical changes to plant structures,

systems or components (SSC). Shutdown

margin will continue to be maintained as

required by plant Technical Specifications to

ensure the reactor will be maintained

sufficiently subcritical to preclude

inadvertent criticality in the shutdown

condition. Shutdown and control rod group

demand position indication is not required to

ensure adequate shutdown margin in modes

3, 4 and 5 and therefore cannot contribute to

the initiation of any accident. The proposed

changes do not change or alter the design

assumptions for the systems or components

used to mitigate the consequences of an

accident, and the initial conditions and

methodologies used in the accident analyses

remain unchanged. Therefore, accident

analyses results are not impacted. Therefore,

the proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not involve any

physical changes to plant structures, systems

or components. The safety functions of the

related structures, systems, or components

are not changed in any manner, nor is the

reliability of any structures, systems, or

components reduced. No new or different

type of equipment will be installed by this

requested change. Therefore, no new failure

modes or potential accident initiators are

introduced. Therefore, the proposed

amendments do not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Shutdown margin will continue to be

maintained in accordance with the

requirements of TS 3/4.1.1. The reactor will

be maintained sufficiently subcritical to

preclude inadvertent criticality in the

shutdown condition. Therefore, the proposed

amendments do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jeffrie J. Keenan,

Nuclear Business Unit—N21, P.O. Box

236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50–

321 and 50–366, Edwin I. Hatch Nuclear

Plant, Units 1 and 2, Appling County,

Georgia

Date of amendment request: June 1,

2000.

Description of amendment request:

The proposed amendments would

revise the vessel pressure and

temperature limit curves that are in the

Technical Specifications.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The changes to the calculational

methodology for the pressure and

temperature (P/T) limits based upon Code

Cases N–640 and N–588 continue to provide

adequate margin in the prevention of a nonductile type fracture of the reactor pressure

vessel (RPV). The code cases were developed

based upon the knowledge gained through

years of industry experience. P/T curves

developed using the allowances of Code

Cases N–640 and N–588 indeed yield more

operating margin. However, the experience

gained in the areas of fracture toughness of

materials and pre-existing undetected defects

show that some of the existing assumptions

used for the calculation of P/T limits are

unnecessarily conservative and unrealistic.

Therefore, providing the allowances of the

subject code cases in developing the P/T

limit curves will continue to provide

adequate protection against nonductile-type

fractures of the RPV.

The evaluation for extending the Unit 1

and Unit 2 P/T limit curves to 54 EFPYs was

performed using the approved methodologies

of 10 CFR 50, Appendix G, and with the

allowances of code cases N–588 and N–640.

The curves generated from these methods

ensure the P/T limits will not be exceeded

during any phase of reactor operation.

Therefore, the probability of occurrence and

the consequences of a previously analyzed

event are not significantly increased. Finally,

the proposed changes will not affect any

other system or piece of equipment designed

for the prevention or mitigation of previously

analyzed events.

Thus, the probability of occurrence and the

consequences of any previously analyzed

event are not significantly increased as the

result of the proposed changes.

2. Do the proposed changes create the

possibility of a new or different type of

accident from any previously evaluated.

Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices 39961

The proposed changes provide more

operating margin in the P/T limit curves for

inservice leakage and hydrostatic pressure

testing, non-nuclear heatup and cooldown,

and criticality, with the benefits being

primarily realizable during the pressure tests.

The revised curves also extend the P/T limit

curves to 54 EFPYs. However, operation in

the ‘‘new’’ regions of the curves have been

analyzed with the new P/T curves providing

adequate protection against a nonductile-type

fracture of the RPV. Otherwise, the proposed

changes do not result in any new or

unanalyzed operation of any system or piece

of equipment important to safety, and as a

result, the possibility of a new type event is

not created.

3. Do the proposed changes involve a

significant reduction in the margin of safety?

As mentioned previously, the revised P/T

curves provide more operating margin and

thus, more operational flexibility than the

current P/T curves. With the increased

operational margin, a reduction in the safety

margin results with respect to the existing

curves. However, the industry experience

since the inception of the P/T limits in 1974

confirms that some of the existing

methodologies used to develop P/T curves

are unrealistic and unnecessarily

conservative. Accordingly, ASME Code Cases

N–640 and N–588 take advantage of the

acquired knowledge by establishing more

realistic methodologies for the development

of P/T curves. Therefore, operational

flexibility is gained and an acceptable margin

of safety to RPV non-ductile type fracture is

maintained.

The extension of the P/T curves to 54

EFPYs was performed per the guidelines of

10 CFR 50, and using code cases N–640 and

N–588 and thus, the margin of safety is not

significantly reduced as the result of the

proposed changes.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Ernest L. Blake,

Jr., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Richard L. Emch,

Jr.

Southern Nuclear Operating Company,

Inc., et al., Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of amendment request: March 3,

2000.

Description of amendment request:

The proposed amendments would

revise technical specification (TS) 3.9.4,

‘‘Containment Penetrations’’, by

allowing the equipment hatch to be

open during core alterations and/or

during movement of irradiated fuel

within the containment.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

No. The proposed changes will allow the

equipment hatch to be open during core

alterations and movement of irradiated fuel

assemblies inside containment. The existing

[Vogtle Electric Generating Plant] VEGP TS

allow the air lock doors to be open during

core alterations and movement of irradiated

fuel assemblies inside containment, and the

dose analyses for a fuel handling accident

inside containment remain bounding for the

case of [an open equipment hatch]. The

proposed changes will not alter the manner

in which fuel is handled or core alterations

are performed. Therefore the proposed

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any previously evaluated?

No. The proposed changes do not create

any new failure modes for any system or

component, nor do they adversely affect

plant operation. No new equipment will be

added and no new limiting single failures

will be created. The plant will continue to be

operated within the envelope of the existing

safety analyses. Therefore, the proposed

changes do not create the possibility of a new

or different kind of accident previously

evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

No. The previously determined

radiological dose consequences for a fuel

handling accident inside containment with

the air lock doors open remain bounding for

the proposed changes. These previously

determined dose consequences were

determined to be well within the limits of 10

CFR 100 and they meet the acceptance

criteria of [Standard Review Plan] SRP

Section 15.7.4 and [General Design Criteria]

GDC 19. Therefore, the proposed changes do

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia

30308–2216.

NRC Section Chief: Richard L. Emch,

Jr.

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Units 1 and 2, Somervell

County, Texas

Date of amendment request: May 17,

2000.

Brief description of amendments: The

proposed amendment would change the

Allowable Values specified in Technical

Specification Table 3.3.5–1 to ensure

that the 6.9 kilovolt (kV) and 480 volt

(V) undervoltage relays initiate the

necessary actions when required. In

addition, some unnecessary limits

would be deleted.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The proposed License Amendment Request

includes more restrictive Allowable Values

for the Preferred offsite source bus

undervoltage function, the Alternate offsite

source bus undervoltage function, the 6.9 kv

Class 1E bus loss of voltage function, the 6.9

kv Class 1E bus degraded voltage function,

and the 480 V Class 1E bus degraded voltage

function. These more restrictive values

assure that all applicable safety analysis

limits are being met. The 480 V low grid

undervoltage relay allowable value is being

lowered to the same as the 480 V degraded

voltage relays which matches its function.

This is a less restrictive value but the value

still assures that all applicable safety analysis

limits are being met. Lowering of the 480 V

low grid undervoltage allowable value will

minimize unnecessary actuations that could

challenge plant systems. Changing the 6.9 kV

and 480 V degraded voltage, 480 V low grid

undervoltage, the 6.9 kV loss of voltage, and

the preferred and alternate bus undervoltage

Allowable Values in the Technical

Specifications has no impact on the

probability of occurrence of any accident

previously evaluated. Because all accident

analyses continue to be met, these changes

do not impact the consequences of any

accident previously evaluated.

Removal of the upper limits for the

preferred and alternate bus undervoltage and

the lower limit for the 6.9 kV Class 1E bus

loss of voltage relays does not impact the

probability of occurrence of any accident

previously evaluated. None of the accident

analyses are affected, therefore, the

consequences of all previously evaluated

accidents remain unchanged.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

None of the changes affect plant hardware

or the operation of plant systems in a way

39962 Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices

that could initiate an accident. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

Response: No.

There were no changes made to any of the

accident analyses or safety analysis limits as

a result of this proposed change. Further, the

proposed change does not affect the

acceptance criteria for any analyzed event.

Removal of the upper limits for the preferred

and alternate source bus undervoltage and

the lower limit for the 6.9 kV Class 1E bus

loss of voltage relays does not change the

margin of safety. Each allowable value, as

revised, assures the safety analysis limits

assumed in the safety analyses as discussed

in Chapter 15 of the FSAR [Final Safety

Analysis Report] is maintained. The margin

of safety established by the Limiting

Conditions for Operation also remains

unchanged. Thus there is no effect on the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: George L. Edgar,

Morgan, Lewis and Bockius, 1800 M

Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: May 25,

2000 (ULNRC–04258).

Description of amendment request:

The proposed amendment would

expand (1) the range of acceptable lift

settings for the pressurizer safety valves

(PSVs), and (2) the tolerance (from +1%

to +2%) of the as-found, measured lift

settings of tested PSVs, to be operable.

The as-left lift settings, following

testing, of the PSVs would not be

changed from the current range of +1%.

The amendment would revise Technical

Specifications (TS) 3.3.2, ‘‘Engineered

Safety Features Actuation System

(ESFAS) Instrumentation,’’ 3.4.10,

‘‘Pressurizer Safety Valves,’’ and 3.4.11,

‘‘Pressurizer Power Operated Relief

Valves (PORVs),’’ of the Callaway TS.

For TS 3.3.2, a new Action H for one or

more trains inoperable would be added,

the note for surveillance requirement

(SR) 3.3.2.14 would be revised to

identify another slave relay that the SR

would be applicable to, and the

automatic PORV actuation would be

added to Table 3.3.2–1, ‘‘Engineered

Safety Features Actuation System

Instrumentation.’’ For TS 3.4.10, the

range of allowable PSV lift settings in

the limiting condition for operation

(LCO) would be expanded from >2460

and <2510 to >2411 and <2509, and SR

3.4.10.1 would be revised to state that

following testing, the lift settings shall

be ‘‘within 1% of 2460 psig’’ instead of

simply ‘‘within 1%.’’ The nominal PSV

lift setting would be changed from 2485

psig to 2460 psig because the maximum

PORV lift setting would not be

increased and the minimum setting

would be reduced 59 psig. For TS

3.4.11, Actions A and B would be

revised to be actions for inoperable

PORVs either solely due to excessive

PORV seat leakage (Action A) or for

reasons other than excessive seat

leakage (Action B), and Action E would

remain an action for two inoperable

PORVs, but would be only for reasons

other than excessive seat leakage. The

licensee also provided corrections to the

Bases of the TSs and the Callaway Final

Safety Analysis Report (FSAR) for the

above changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The pressurizer safety valves (PSVs), in

conjunction with the Reactor Trip System

(RTS), provide overpressure protection for

the Reactor Coolant System (RCS). The PSV

[lift] setpoint is established to maintain the

RCS pressure below 110% of the system

design pressure. The proposed change in the

minimum allowable PSV setpoint could

result in a transient being terminated at a

pressure that is lower than that assumed in

the transient’s analysis. However, the

primary system pressure boundary is not

challenged by the minimum allowable PSV

setpoint. Since the maximum allowable PSV

setpoint is unaffected by the proposed

change (other than from round-off, as

discussed previously [in the application,

from 2510 to 2509 psig]), the primary system

pressure boundary is not challenged by the

maximum allowable PSV setpoint.

With a nominal setpoint of 2460 psig and

a [as-found] +2% setpoint tolerance, the PSV

actuation setpoint could potentially open at

pressures as low as 2410 psig (rounded up in

revised LCO 3.4.10 to 2411 psig). This lower

PSV actuation setpoint will reduce the

margin between the pressurizer PORV and

PSV actuation setpoint from 125 psi to 75

psi. A 75 psi margin is considered adequate

and should not challenge the PSVs on

Condition I transients.

The majority of the Callaway PRA

[probabilistic risk assessment] event trees

question the capability of the PORVs to open

for RCS cooldown and depressurization or

for feed and bleed cooling. Some event trees

question the capability of the PORVs to

reclose to terminate RCS depressurization

and coolant inventory loss. The transientinduced ATWS [anticipated transient

without scram] event trees question the

capability of the PSVs to reclose after

opening for these high pressure transients.

The maximum allowable PSV setpoint is

essentially unchanged; therefore, the

proposed change will not adversely impact

the probability of the PSVs failing open.

Upgrading the automatic PORV actuation

circuitry to fully Class 1E, and revising the

Technical Specification operability and

surveillance requirements to demonstrate the

operability of the automatic PORV actuation

circuitry, will enhance valve reliability and

assure compliance with NRC Generic Letter

90–06. However, it has been determined that

this plant modification increase the

probability that the PORVs will inadvertently

open and remain open if multiple transmitter

failures are postulated. With the new safety

grade PORV 2/4 [two out of four] opening

actuation logic, two failed high pressurizer

pressure channels would result in

inadvertent opening of both PORVs and the

PORVs would remain open until remotemanually closed. Since two of the four

channels available to reclose the PORVs are

assumed to have failed high, and since

closure of the PORVs would require a 3/4

logic to close after the modification is

implemented, there would be no signal to

close the PORVs on a low pressurizer

pressure signal. With the current opening

logic, a single failed high pressurizer

pressure channel would result in opening

one PORV. However, the current 2/4 closure

logic would reclose that PORV when

pressurizer pressure drops below

approximately 2200 psia. With the current

control logic, three failed high pressurizer

pressure channels (3/4) are required for both

PORVs to inadvertently open and remain

open. However, the consequences of both

PORVs inadvertently opening and remaining

open are bounded by the analysis in FSAR

Section 15.6.1, ‘‘Inadvertent Opening of a

Pressurizer Safety or Relief Valve.’’ Since a

pressurizer safety valve is sized to relieve

approximately twice the steam flow rate of a

pressurizer PORV, and will therefore allow a

much more rapid depressurization upon

opening, the analysis in Section 15.6.1

examines the accidental depressurization of

the RCS associated with an inadvertent

opening of a pressurizer safety valve. While

there is no way to isolate a stuck-open

pressurizer safety valve, two open PORVs can

be remote-manually isolated by either closing

the PORVs or the PORV block valves. Since

there is a small impact due to multiple

channel failures resulting in an increase in

the probability of both PORVs inadvertently

opening and remaining open, it is concluded

that the proposed activity increases the

probability of occurrence of an accident

previously evaluated in the FSAR. However,

multiple failures are required for this

malfunction and failure modes that result in

multiple channels failing high are highly

unlikely. Therefore, this increase in the

probability that the PORVs will inadvertently

open and remain open is considered to be

insignificant.

Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices 39963

All evaluations performed for overpressure

transients conservatively assume the upper

limit of the PSV tolerance as the pressure to

which the RCS is subjected. It has been

determined that the design transients are not

adversely affected because the limiting

transients are not sensitive to the pressure

tolerance change. Although the lower PSV

setpoint would result in a lower PSV relief

flow rate, the slightly lower valve flow rate

would be more than compensated for by the

reduced valve opening pressure. The change

to the PSV setpoint and setpoint tolerance

does not change the conclusions of the

existing thermal-hydraulic and stress

analyses for the pressurizer safety and relief

system. The design function of the valves is

not being changed and the conclusions

documented in the NRC Safety Evaluation of

Callaway’s response to NUREG–0737 Item

II.D.1[‘‘Performance Testing of the

Pressurizer Power-Operated Relief Valve,’’]

(dated September 10, 1987) are unchanged

(see also FSAR Section 18.2.5). The PORVs

and associated discharge piping can

accommodate water relief.

Overall protection system performance will

remain within the assumptions of the

previously performed accident analyses since

the only hardware changes are associated

with making the automatic PORV actuation

circuitry fully Class 1E. The RTS and

Engineered Safety Features Actuation System

(ESFAS) protection systems will continue to

function in a manner consistent with the

plant design basis. The automatic PORV

actuation circuitry modification will be

performed in such a manner that all design,

material, and construction standards that

were applicable to safety-related systems

prior to the change are maintained.

The proposed change will not affect the

probability of any event initiators nor will

the proposed change negatively affect the

ability of any safety-related equipment to

perform its intended function. Changing the

PSV lift setting does not change the

probability that an event will occur which

will result in the PSV opening. There will be

no degradation in the performance of safetyrelated equipment assumed to function

during an accident situation. There will be

no change to normal plant operating

parameters.

Since the FSAR Chapter 15 LOCA [loss-ofcoolant accident], SGTR [steam generator

tube rupture] and MSLB [main steam line

break] analyses all result in decreasing RCS

pressure and do not challenge the PSV

opening pressure, none of these events are

affected by the proposed change to the PSV

nominal setpoint and the allowable setpoint

tolerance. Timely operator actions will be

taken to preclude water relief through the

PSVs during an Inadvertent ECCS

[emergency core cooling system] Actuation at

Power event. Water relief from the PORVs for

the latter event would result in a larger

discharge of RCS inventory than currently

analyzed, wherein operator action is assumed

to terminate safety injection within 10

minutes prior to the pressurizer filling.

However, FSAR Figure 15.5–3 in Attachment

5 [to the application] demonstrates that DNB

[departure from nucleate boiling] is not a

concern, there will be no fuel failures

associated with this event, and RCS

inventory will be directed to the pressurizer

relief tank located inside containment.

Therefore, there will be no impact on offsite

radiological consequences. None of the other

non-LOCA transients are adversely affected

by the proposed change. Since none of the

other FSAR Chapter 15 events are adversely

affected, the radiological consequences of

those events are not adversely affected.

In the Westinghouse reanalysis of the

Inadvertent ECCS Actuation at Power event,

the minimum PSV opening setpoint serves as

a limit to demonstrate the acceptability of the

assumed operator action times to assure that

the PSVs will not be required to operate

while the pressurizer is water solid. A lower

PSV opening setpoint could potentially

require earlier operator actions to prevent

water relief through the PSVs. Simulator

exercises for the Inadvertent ECCS Actuation

at Power event were performed on the

Callaway training simulator on August 10,

1999 to determine the times required for the

control room operators to stop the NCP

[normal charging pump] and unblock the

PORVs and assure their availability for

automatic pressure relief. In all cases, the

NCP was stopped within four (4) minutes

and the PORVs were unblocked and available

for automatic pressure relief within seven (7)

minutes. The reanalysis in Attachment 5 [to

the application] conservatively credits

operator actions from the main control room

to stop the NCP in six (6) minutes and to

unblock the PORVs and assure their

availability for automatic pressure relief in

nine (9) minutes. These times include all

process and instrumentation delays. The

revised FSAR Figure 15.5–2 shows that if

operator actions are taken within these time

frames to terminate NCP flow and to assure

at least one PORV is available for automatic

pressure relief, water relief through the PSVs

is precluded. Procedure changes and periodic

operator requalification training will provide

assurance that these operator actions can be

performed within the assumed time

constraints.

Based on the above discussions, the

proposed change will not involve a

significant increase in the probability of

occurrence or the consequences of an

accidently previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The nominal setpoint for the PSVs will be

lowered by 1% from 2485 psig to 2460 psig.

The allowable setpoint tolerance will be

increased from +1% to +2%. The combined

effect of these changes results in a 2%

decrease in the minimum acceptable PSV

[lift] setpoint from 2460 psig to 2411 psig.

The change in the PSV setpoint and in the

tolerance of the setpoint does not change

their ability to open on demand. The

maximum acceptable PSV setpoint is

unaffected by this proposed change, other

than round-off as discussed previously. Since

the FSAR accident analyses do not rely on

the automatic actuation of non-safety related

control grade systems or components for

accident mitigation, a plant modification will

make the automatic pressurizer PORV

pressure relief circuitry fully Class 1E.

The proposed change to the PSV nominal

setpoint and the allowable setpoint tolerance

will not prevent the PSVs from performing

their RCS overpressurization protection

function. Additionally, the proposed change

does not affect the ability of any other safetyrelated equipment to perform its safety

function.

The only hardware changes are associated

with making the automatic PORV actuation

circuitry fully Class 1E. The RTS and

Engineered Safety Feature Actuation System

(ESFAS) protection systems will continue to

function in a manner consistent with the

plant design basis. The automatic PORV

actuation circuitry modification will be

performed in such a manner that all design,

material, and construction standards that

were applicable to safety-related systems

prior to the change are maintained. While the

possibility that the PORVs fail to control RCS

pressure, that at least one PORV fails to open,

and that the operator fails to open the block

valve and assure the PORV(s) are available

for automatic pressure relief within the

required time frame are all malfunctions of

a different type than currently analyzed in

the FSAR, they do not create different

accident types. The Class 1E upgrade and

changes to Emergency Operating Procedure

E–0 will provide assurance that the

reanalysis presented in Attachment 5 [to the

application] will bound the results of this

event which, in turn, is also bounded by the

results presented in FSAR Section 15.6.1 for

an inadvertent PSV opening.

There are no other changes in the method

by which any safety-related plant system

performs its safety function. The change will

not affect the normal method of plant

operation.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The PSVs, in conjunction with the RTS,

provide overpressure protection for the RCS.

The change in the upper limit of the PSV

tolerance from +1% to +2%, with a reduction

in the nominal setpoint from 2485 psig to

2460 psig, does not challenge the upper limit

of overpressure protection. The maximum

opening pressure setpoint is unchanged

(other than a conservative round-off), and

therefore, does not impact analyses

performed for overpressure transients. The

change to the PSV setpoint and setpoint

tolerance does not change the conclusions of

the existing thermal-hydraulic and stress

analyses for the pressurizer safety and relief

system. For all non-LOCA events, the above

evaluations support the change in the PSV

setpoint and setpoint tolerance from 2485

psig +1% to 2460 psig +2%. The change in

the PSV setpoint and setpoint tolerance also

has no effect on the RTS and ESFAS trip

setpoints.

The Bases for Technical Specification

3.4.10 states the following in the Background

section:

‘‘The safety valves are designed to prevent

the system pressure from exceeding the

system Safety Limit (SL), 2735 psig, which is

110% of the design pressure * * * The relief

39964 Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices

capacity for each valve, 420,000 lb/hr at 2485

psig plus 3% accumulation, is based on

postulated overpressure transient conditions

resulting from a complete loss of steam flow

to the turbine. This event results in the

maximum surge rate into the pressurizer

* * * .’’

The locked RCP [reactor coolant pump]

rotor and loss of external electrical load/

turbine trip transient analyses assume PSV

actuation at 2550 psia. This value is

conservatively based on a nominal PSV

setpoint of 2500 psia plus a 1% setpoint

tolerance and a 1% setpoint shift (due to the

presence of the water seal). The maximum

allowable PSV setpoint of 2509 psig is

unaffected by the proposed change, other

than a conservative round-off discussed

previously. At a pressure of 2509 psig, the

minimum relief capacity of the safety valves

would be in excess of 420,000 lb/hr.

However, the safety analyses for

overpressurization events conservatively

assume a 420,000 lb/hr minimum design

relief capacity for the PSVs.

The proposed change does not affect the

acceptance criteria for any other analyzed

event nor is there a change to any other

Safety Analysis Limit (SAL). The acceptance

criteria for the Inadvertent ECCS Actuation at

Power event will remain the same as

currently analyzed; however, operator action

and automatic PORV actuation will be relief

upon to demonstrate compliance with that

event’s acceptance criteria.

There will be no effect on the manner in

which safety limits or limiting safety system

settings are determined nor will there be any

effect on those plant systems necessary to

assure the accomplishment of protection

functions. There will be no impact on the

overpower limit, DNBR limits, FQ, F∆H,

LOCA PCT [peak cladding temperature], peak

local power density, or any other margin of

safety. The radiological dose consequence

acceptance criteria listed in the [NRC]

Standard Review Plan continue to be met.

Therefore, the proposed change does not

involve a significant reduction in any margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: John O’Neill,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW, Washington, DC

20037.

NRC Section Chief: Stephen Dembek.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

FirstEnergy Nuclear Operating

Company, Docket No. 50–440, Perry

Nuclear Power Plant, Unit 1, Lake

County, Ohio

Date of amendment request: June 1,

2000.

Description of amendment request:

The proposed amendment would permit

changes to the Perry Nuclear Power

Plant Updated Safety Analysis Report

(USAR) to incorporate descriptions (in

the form of text, tables, and drawings)

of modifications to the Emergency

Service Water (ESW) alternate intake

sluice gate. The modifications will

include (1) installation of a safetyrelated Class 1E selector switch that will

be used to disable the automatic

opening function of the sluice gate

during warm weather and (2)

installation of a non-safety inflatable

sealing device on the gates between the

ESW forebay and the alternate intake

tunnel. The modifications are designed

to increase overall reliability of the ESW

system and to eliminate undesired

operation of the ESW pumps.

Date of publication of individual

notice in Federal Register: June 14, 2000

(65 FR 37414).

Expiration date of individual notice:

July 14, 2000.

Attorney for licensee: Mary E.

O’Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street,

Akron, OH 44308.

NRC Section Chief: Anthony J.

Mendiola.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and

electronically from the ADAMS Public

Library component on the NRC Web

site, http://www.nrc.gov (the Electronic

Reading Room).

AmerGen Energy Company, LLC, Docket

No. 50–461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of application for amendment:

April 24, 2000.

Brief description of amendment: The

amendment allowed a one-time

extension of some Technical

Specification surveillance intervals due

to elimination of a planned midcycle

outage. The surveillances would be

extended to no later than November 30,

2000.

Date of issuance: June 12, 2000.

Effective date: Immediately, to be

implemented within 30 days.

Amendment No.: 129.

Facility Operating License No. NPF–

62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 8, 2000 (65 FR 26642).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 12, 2000.

No significant hazards consideration

comments received: No.

Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices 39965

Commonwealth Edison Company,

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of application for amendments:

March 15, 2000.

Brief description of amendments: The

amendments revised the Technical

Specifications to permit plant operation

with an ultimate heat sink temperature

of 100 °F.

Date of issuance: June 13, 2000.

Effective date: Immediately as of the

date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 107 and 107.

Facility Operating License Nos. NPF–

72 and NPF–77: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 3, 2000 (65 FR 25763).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 13, 2000.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

July 27, 1999, as supplemented by

letters dated October 7, 1999, and May

31, 2000.

Brief description of amendments: The

amendments revised the Technical

Specifications by adding a surveillance

requirement to verify the Keowee out-oftolerance logic trips and blocks closure

of the appropriate overhead or

underground power path breakers.

Date of Issuance: June 6, 2000.

Effective date: As of the date of

issuance and shall be implemented by

November 30, 2000.

Amendment Nos.: 312, 312 and 312.

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: Amendments

revised the TS.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR 46429).

The supplements dated October 7, 1999,

and May 31, 2000, provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 6, 2000.

No significant hazards consideration

comments received: No.

Portland General Electric Company, et

al., Docket No. 50–344, Trojan Nuclear

Plant, Columbia County, Oregon

Date of application for amendment:

November 16, 1999.

Brief description of amendment: The

amendment revised the Permanently

Defueled Technical Specifications by

removing Figure 4.1–1, ‘‘Site and

Exclusion Area Boundaries,’’ and

incorporating the applicable portions of

this figure in the Trojan Defueled Safety

Analysis Report. Other associated

administrative changes resulting from

the deletion of Figure 4.1–1, as well as

an administrative change to the table of

contents, were also made.

Date of issuance: May 31, 2000.

Effective date: May 31, 2000.

Amendment No.: 204.

Facility Operating License No. NPF–1:

The amendment changes the

Permanently Defueled Technical

Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4289).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 31, 2000.

No significant hazards consideration

comments received: No.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

February 9, 2000.

Brief description of amendment: This

amendment revises Technical

Specification (TS) Limiting Condition

for Operation 3.8.2.1 to add two new

Action Statements for operating

conditions where a Class 1E battery’s

electrolyte temperature is below the

minimum limit specified in TS

Surveillance Requirement 4.8.2.1.b.3.

Date of issuance: June 9, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 60 days.

Amendment No.: 127.

Facility Operating License No. NPF–

57: This amendment revised the TSs.

Date of initial notice in Federal

Register: March 8, 2000 (65 FR 12294).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 9, 2000.

No significant hazards consideration

comments received: No.

Tennessee Valley Authority, Docket No.

50–390, Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of application for amendment:

September 28, 1999, as supplemented

March 17, 2000.

Brief description of amendment:

Revised Technical Specifications

definitions for Engineered Safety

Feature Response Time and Reactor

Trip System Response Time, to provide

for verification of response time for

selected components, provided that the

components and the methodology for

verification have been previously

reviewed and approved by the NRC.

Date of issuance: June 13, 2000.

Effective date: June 13, 2000.

Amendment No.: 24.

Facility Operating License No. NPF–

90: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: October 20, 1999 (64 FR

56534). The March 17, 2000, submittal

provided clarifying information that did

not change the scope of the original

request or change the initial proposed

no significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 13, 2000.

No significant hazards consideration

comments received: No.

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Unit Nos. 1 and 2, Somervell

County, Texas

Date of amendment request:

November 8, 1999, as supplemented by

letters dated April 13, and May 30,

2000.

Brief description of amendments: The

amendments change Technical

Specification 5.5.11, ‘‘Ventilation Filter

Testing Program (VFTP),’’ to include the

requirement for laboratory testing of

Engineered Safety Feature (ESF)

Ventilation System charcoal samples

per American Society for Testing and

Materials D3803–1989 and the

application of a safety factor of 2.0 to

the charcoal filter efficiency assumed in

the plant design-basis dose analyses.

The license amendments also extend the

implementation date for License

Amendment 74, currently June 30, 2000,

to December 31, 2000.

Date of issuance: June 12, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 78 and 78.

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 29, 1999 (64 FR

73101). The April 13, and May 30, 2000,

letters provided clarifying information

that did not change the scope of the

November 8, 1999, application nor the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 12, 2000.

39966 Federal Register / Vol. 65, No. 125 / Wednesday, June 28, 2000 / Notices

1The radioactive materials, sometimes referred to

as agreement materials, are: (a) Byproduct materials

as defined in Section 11e.(1) of the Act; (b)

byproduct materials as defined in Section 11e.(2) of

the Act; (c) source materials as defined in Section

11z. of the Act; and (d) special nuclear materials as

defined in Section 11a. of the Act, restricted to

quantities not sufficient to form a critical mass.

No significant hazards consideration

comments received: No.

Viacom Inc., Docket No. 50–22, Test

Reactor, Waltz Mill, Pennsylvania

Date of application for amendment:

February 14, 2000 supplemented on

March 8 and 25, 2000.

Brief description of amendment: This

amendment changes the license to

reflect the transfer of the licensee for the

Test Reactor at Waltz Mill from the CBS

Corporation to Viacom Inc.

Date of issuance: May 31, 2000.

Effective Date: May 4, 2000.

Amendment No.: 12.

Facility License No. TR–2: This

amendment changes the license.

Date of Initial notice in Federal

Register: February 29, 2000 (65 FR

10841).

The Commission has issued a Safety

Evaluation for this amendment dated

April 13, 2000.

No significant hazards consideration

comments received: No.

Local Public Document: N/A.

Dated at Rockville, Maryland, this 21st day

of June 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–16193 Filed 6–27–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

State of Oklahoma: NRC Staff

Assessment of a Proposed Agreement

Between the Nuclear Regulatory

Commission and the State of

Oklahoma

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of a proposed agreement

with the State of Oklahoma.

SUMMARY: This notice is announcing

that the Nuclear Regulatory Commission

(NRC) has received a request from

Governor Frank Keating of Oklahoma

that the NRC consider entering into an

Agreement with the State as authorized

by Section 274 of the Atomic Energy Act

of 1954, as amended (Act). Section 274

of the Act contains provisions for the

Commission to enter into agreements

with the Governor of any State

providing for the discontinuance of the

regulatory authority of the Commission.

Under the proposed Agreement,

submitted December 28, 1999, the

Commission would discontinue and

Oklahoma would take over portions of

the Commission’s regulatory authority

over radioactive material covered under

the Act within the State of Oklahoma.

In accordance with 10 CFR 150.10,

persons, who possess or use certain

radioactive materials in Oklahoma,

would be released (exempted) from

portions of the Commission’s regulatory

authority under the proposed

Agreement. The Act requires that NRC

publish those exemptions. Notice is

hereby given that the pertinent

exemptions have been previously

published in the Federal Register and

are codified in the Commission’s

regulations as 10 CFR Part 150. NRC is

publishing the proposed Agreement for

public comment, as required by the Act.

NRC is also publishing the summary of

an assessment conducted by the NRC

staff of the proposed Oklahoma

byproduct material regulatory program.

Comments are invited on (a) the

proposed Agreement, especially its

effect on public health and safety, and

(b) the NRC staff assessment.

DATES: The comment period expires July

7, 2000. Comments received after this

date will be considered if it is practical

to do so, but the Commission cannot

assure consideration of comments

received after the expiration date.

ADDRESSES: Written comments may be

submitted to Mr. David L. Meyer, Chief,

Rules and Directives Branch, Division of

Administrative Services, Office of

Administration, Washington, DC 20555–

0001. Copies of comments received by

NRC may be examined at the NRC

Public Document Room, 2120 L Street,

NW. (Lower Level), Washington, DC.

Copies of the proposed Agreement,

copies of the request for an Agreement

by the Governor of Oklahoma including

all information and documentation

submitted in support of the request, and

copies of the full text of the NRC staff

assessment are also available for public

inspection in the NRC’s Public

Document Room.

FOR FURTHER INFORMATION CONTACT:

Patricia M. Larkins, Office of State and

Tribal Programs, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001. Telephone (301) 415–

2309 or e-mail pml@nrc.gov.

SUPPLEMENTARY INFORMATION: Since

Section 274 of the Act was added in

1959, the Commission has entered into

Agreements with 31 States. The

Agreement States currently regulate

approximately 16,000 agreement

material licenses, while NRC regulates

approximately 5800 licenses. Under the

proposed Agreement, approximately

220 NRC licenses will transfer to

Oklahoma. NRC periodically reviews

the performance of the Agreement States

to assure compliance with the

provisions of Section 274. Section 274e

requires that the terms of the proposed

Agreement be published in the Federal

Register for public comment once each

week for four consecutive weeks. This

notice is being published in fulfillment

of the requirement.

I. Background

(a) Section 274d of the Act provides

the mechanism for a State to assume

regulatory authority, from the NRC, over

certain radioactive materials 1 and

activities that involve use of the

materials. In a letter dated December 28,

1999, Governor Keating certified that

the State of Oklahoma has a program for

the control of radiation hazards that is

adequate to protect public health and

safety within Oklahoma for the

materials and activities specified in the

proposed Agreement, and that the State

desires to assume regulatory

responsibility for these materials and

activities. Included with the letter was

the text of the proposed Agreement,

which is included as Appendix A to this

notice.

The radioactive material and activities

(which together are usually referred to

as the ‘‘categories of material’’) which

the State of Oklahoma requests

authority over are: (1) The possession

and use of byproduct materials as

defined in Section 11e.(1) of the Act; (2)

the possession and use of special

nuclear material in quantities not

sufficient to form a critical mass; (3) the

regulation of the land disposal of

byproduct source or special nuclear

material received from other persons;

and (4) source material used to take

advantage of its density and high mass

properties where the use of the

specifically licensed source material is

subordinate to the primary specifically

licensed use of either 11e.(1) byproduct

material or special nuclear material, as

provided for in regulations or orders of

the Commission.

(b) The proposed Agreement contains

articles that:

—Specify the materials and activities

over which authority is transferred;

—Specify the activities over which the

Commission will retain regulatory

authority;

—Continue the authority of the

Commission to safeguard nuclear

materials and restricted data;