Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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 April 12, 1997, through April 25, 1997. The

last biweekly notice was published on April 23, 1997 (62 FR 19825).

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.

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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 June 6, 1997, 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 at the local public

document room for the particular facility involved. 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 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. Where petitions are filed during the last 10 days of

the notice period, it is requested that the petitioner promptly so

inform the Commission by a toll-free telephone call to Western Union at

1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to (Project Director): petitioner's name

and telephone number, date petition was mailed, plant name, and

publication date and page number of this Federal Register notice. 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 at the local public document room for

the particular facility involved.

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley

Power Station, Unit No. 1, Shippingport, Pennsylvania

Date of amendment request: March 10, 1997

Description of amendment request: The proposed amendment would

modify the Technical Specifications (TSs) by reducing the reactor

coolant system (RCS) specific activity limits in accordance with

Generic Letter 95-05. The definition of DOSE EQUIVALENT I-131 would be

replaced with the Improved Standard TS definition wording in the first

sentence and an

[[Page 24986]]

equation added based on dose conversion factors derived from

International Commission on Radiation Protection (ICRP) ICRP-30. TS

3.4.8, Specific Activity, would be revised by reducing the DOSE

EQUIVALENT I-131 limit from 1.0 [micro]Ci[curies]/gram to 0.35

[micro]Ci[curies]/gram. Item 4.a in TS Table 4.4-12, Primary Coolant

Specific Activity Sample and Analysis Program, TS Figure 3.4-1, and the

Bases for TS 3/4.4.8 would be modified to reflect the reduced DOSE

EQUIVALENT I-131 limit.

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 proposed change reduces the reactor coolant system (RCS)

specific activity limits of Specification 3.4.8 from 1.0 [micro]Ci/

gram to 0.35 [micro]Ci/gram and lowers the graph in Figure 3.4-1 by

39 [micro]/Ci gram following the guidance provided in Generic Letter

(GL) 95-05. This reduces the RCS activity allowed to leak to the

secondary side when the plant is operating so that additional margin

is available to support a higher allowable accident-induced leakage

value as justified by analysis.

The proposed changes to Specification 3.4.8 and the definition

of DOSE EQUIVALENT I-131 ensure these requirements are consistent

the latest analyses.

These changes implement the more restrictive RCS activity limits

in accordance with applicable analyses and GL 95-05 to ensure the

regulations are satisfied. Therefore, these changes do not involve a

significant 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 alter the configuration of the

plant or affect the operation with the reduced specific activity

limit. By reducing the specific activity limit, the limit would be

reached sooner to initiate evaluation of the out of limit condition.

The proposed changes will not result in any additional challenges to

the main steam system or the reactor coolant system pressure

boundary. Consequently, no new failure modes are introduced as a

result of the proposed changes. As a result, the main steam line

break, steam generator tube rupture and loss of coolant accident

analyses remain bounding. Therefore, the proposed change will not

create the possibility of a new or different kind of accident from

any accident previously evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed change reduces the RCS specific activity limit to

0.35 [micro]Ci/gram along with lowering the Figure 3.4-1 limits by

39 [micro]Ci/gram. Reduction of the RCS specific activity limits

allows an increase in the limit for the projected SG [steam

generator] leakage following SG tube inspection and repair in

accordance with the voltage-based SG tube alternate repair criteria

(ARC) incorporated by Amendment No. 198. This follows the guidance

provided in GL 95-05 and effectively takes margin available in the

specific activity limits and applies it to the projected SG leakage

for the ARC. This has been determined to be an acceptable means for

accepting higher projected leakage rates while still meeting the

applicable limits of 10 CFR [Part] 100 and GDC [General Design

Criterion] 19 with respect to offsite and control room doses.

The capability for monitoring the specific activity and

complying with the required actions remains unchanged. In addition,

there is no resultant change in dose consequences. Therefore, the

proposed change does 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.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037

NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: March 14, 1997

Description of amendment request: The proposed amendment would

relocate the following administrative control technical specifications

(TSs) from the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1

and BVPS-2) TSs to the quality assurance program description, which is

presented in Section 17.2 of the BVPS-2 Updated Final Safety Analysis

Report (UFSAR). Section 17.2 of the BVPS-2 UFSAR contains the quality

assurance program description for both BVPS-1 and BVPS-2. The licensee

stated that the proposed changes are based on NRC Administrative Letter

95-06, ``Relocation of Technical Specification Administrative Controls

Related to Quality Assurance.''

BVPS-2 TS 6.2.3 (Independent Safety Evaluation Group)

BVPS-1 and BVPS-2 TS 6.5.1 (Onsite Safety Committee)

BVPS-1 and BVPS-2 TS 6.5.2 (Offsite Review Committee)

BVPS-1 and BVPS-2 TS 6.8.2 (Procedures, Review and Approval)

BVPS-1 and BVPS-2 TS 6.8.3 (Temporary Procedure Changes, Review and

Approval)

BVPS-1 and BVPS-2 TS 6.10.1 (Records Retention, At least 5 years)

BVPS-1 and BVPS-2 TS 6.10.2 (Records Retention, Duration of

Operating License)

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?

This proposed change would relocate technical specification

administrative controls to the quality assurance program

description. Adequate controls are provided by the established

quality assurance program change process in 10 CFR 50.54(a).

The provisions of Technical Specification 6.2.3.2 which states

that: ``The ISEG [Independent Safety Evaluation Group] shall be

composed of at least five, dedicated, full-time engineers located on

site,'' would be omitted from the provisions relocated to the

quality assurance program description. Since no system, component or

operational procedure changes are involved, and the ISEG function

will continue to be implemented, the change can have no effect on

safe operation of the plant.

The likelihood that an accident will occur is not increased by

this proposed technical specification change which involves

administrative controls. No systems, equipment, or components are

affected by the proposed change. Thus, the consequences of a

malfunction of equipment important to safety previously evaluated in

the Updated Final Safety Analysis Report (UFSAR) are not increased

by this change.

Relocation of technical specification provisions and related

changes do not affect possible initiating events for accidents

previously evaluated or any system functional requirement. The

proposed changes have no impact on accident initiators or plant

equipment, and do not affect the probabilities or consequences of an

accident.

Therefore, the proposed changes will not involve a significant

increase in the probability or consequences of a previously

evaluated accident.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed relocation of technical specification provisions to

the quality assurance program description and related changes do not

involve changes to the physical plant or operations. Since the

proposed changes to administrative controls do not affect equipment

or its operation, they cannot contribute to accident initiation and

[[Page 24987]]

cannot produce a new accident scenario or a new type of equipment

malfunction.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes are administrative in nature and do not

directly affect plant equipment or operation. Safety limits and

limiting safety system settings are not affected by this proposed

change. The proposed changes do not affect the UFSAR design bases,

accident assumptions, or technical specification bases. In addition,

the proposed changes do not affect release limits, monitoring

equipment or practices.

Therefore, the proposed changes would 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.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: John F. Stolz

Entergy Operations Inc., Docket No. 50-382, Waterford Steam

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: April 11, 1997

Description of amendment request: The proposed amendment modifies

Technical Specification (TS) 3.3.3.7.3 and Surveillance Requirement

4.3.3.7.3 for the broad range gas detection system at Waterford Steam

Electric Station, Unit 3. The proposed change also includes changes in

TS Basis 3/4.3.3.7.3 to support the 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. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

The broad range gas detection system has no effect on the

accidents analyzed in chapter 15 of the Final Safety Analysis

Report. It's only effect is on habitability of the control room,

which will be enhanced by installation of the new monitoring system

and this change to the Technical Specifications. Analysis has shown

that the impact on operator incapacitation and subsequent core

damage risk of this background check is negligible.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different type of

accident from any accident previously evaluated?

Response: No.

The proposed Technical Specification change in itself does not

change the design or configuration of the plant. The new system for

broad range toxic gas monitoring performs the same function as the

old system, but it accomplishes this with a more sophisticated

system that increases reliability.

Therefore, the proposed change will not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

The broad range gas detection system has no effect on a margin

of safety as defined by Section 2 of the Technical Specifications.

It's only effect is on habitability of the control room, which will

be enhanced by installation of the new monitoring system and this

change to the Technical Specifications.

Therefore, the proposed change 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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

Northeast Nuclear Energy Company, et al., Docket No. 50-336,

Millstone Nuclear Power Station, Unit No. 2, New London County,

Connecticut

Date of amendment request: April 10, 1997

Description of amendment request: The proposed changes would modify

the Technical Specifications (TSs) for the Enclosure Building. The

Enclosure Building is a limited-leakage, steel-framed structure that

completely surrounds the containment. It is designed and constructed to

ensure that any leakage of radioactive materials to the environment

would not exceed an acceptable upper limit in the event of a design

basis loss-of-coolant accident or movement of loads over the spent fuel

pool. A slight negative pressure is maintained by the Enclosure

Building Filtration System and the system exhausts the filtered air

through charcoal and high-efficiency particulate air (HEPA) filters.

Specifically, the proposed changes would relocate the surveillance

requirement for attaining a negative pressure in the Enclosure Building

from TS 3.6.5.1 ``Enclosure Building Filtration System,'' to TS

3.6.5.2, ``Enclosure Building Integrity.'' TS 3.6.5.2 would also be

changed to address operability, which includes integrity requirements,

and the Definition 1.25, ``Enclosure Building Integrity,'' would be

deleted. TS 4.6.5.2, ``Surveillance Requirements,'' would be modified

to require each access opening in the Enclosure Building to be closed

instead of the current requirement to close each door (some access

openings have two doors in series) in each access opening. This TS

would also be renumbered as 4.6.5.2.1.

In addition, editorial changes are proposed for consistency and the

index pages would be updated to reflect the proposed changes. The TS

Bases would also be updated to reflect the proposed changes including

the need to maintain the integrity of the Enclosure Building and to

support previously approved laboratory testing requirements for

charcoal filter sample testing.

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. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to Technical Specifications 3.6.5.1 and

3.6.5.2, relocation of Surveillance Requirement 4.6.5.1.d.3 to

Specification 3.6.5.2, changes to Bases Sections 3.6.5.1 and

3.6.5.2, and deletion of Definition 1.25 will resolve the conflict

that currently exists between Specifications 3.6.5.1 and 3.6.5.2.

Specifically, the requirement to establish and maintain a negative

pressure in the Enclosure Building boundary included in

Specification 3.6.5.1 belongs in Specification 3.6.5.2. In the event

Enclosure Building operability is not maintained in Modes 1-4, the

Action Statement for LCO [limiting condition for operation] 3.6.5.2

requires that Enclosure Building operability must be restored within

24 hours. Twenty-four hours is a reasonable completion time

considering the limited leakage design of containment and the low

[[Page 24988]]

probability of a DBA [design-basis accident] occurring during this

time period. Therefore, it is considered that there exists no loss

of safety function. The

proposed changes do no modify the LCO or surveillance acceptance

criterion, nor do they change the frequency of the surveillances.

The proposed changes do not involve any physical changes to the

plant, do not alter the way any structure, system, or component

functions. Therefore, the structures, systems, or components will

perform their intended function when called upon. (The redundancy of

the double doors has not been credited in the radiological dose

calculations for any Design Basis Accident.) Additionally, the

proposed changes are consistent with the new, improved Standard

Technical Specifications for Combustion Engineering plants (NUREG-

1432).

The editorial changes to Technical Specifications 3.6.5.1,

3.6.5.2, and 3.9.15 do not change any technical aspect of these

specifications. Therefore the proposed changes do not affect the

probability of any previously evaluated accident.

Based on the above, the proposed changes do 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.

The proposed changes do not make any physical or operational

changes to existing plant structures, systems, or components. The

proposed changes do not introduce any new failure modes. The

proposed changes simply resolve a conflict which currently exits

between Specifications 3.6.5.1 and 3.6.5.2. Thus, the proposed

changes do not create the possibility of a new or different kind of

accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed changes do not have any adverse impact on the

accident analyses. Also, the proposed changes resolve a conflict

which currently exists between Specifications 3.6.5.1 and 3.6.5.2.

The structures, systems, or components covered under Specifications

3.6.5.1 and 3.6.5.2 will perform their intended safety function when

called upon.

Based on the above, there is no 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.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270

NRC Deputy Director: Phillip F. McKee

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power

Station, Units Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: March 31, 1997

Description of amendment request: The proposed change revises the

Peach Bottom Atomic Power Station, Units 2 and 3 technical

specifications to extend the surveillance interval for calibration of

Average Power Range Monitor (APRM) flow bias instrumentation from 18

months to 24 months.

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

because the accidents previously evaluated take credit only for the

clamped 120% high neutron flux scram setpoint. Credit is not taken

for the flow biased APRM scram setpoint. Failure or inaccuracy of

the flow biased feature of the APRM scram setpoint will in no way

affect the clamped high flux scram setpoint. The 120% high flux

scram setpoint is derived internal to the APRM circuitry and

calibrated separately as part of the APRM trip circuitry. The APRM

clamped high flux scram setpoint is not being impacted by the

proposed changes and will be automatically enforced regardless of

the status or accuracy of the APRM flow bias circuitry.

Because there is no impact on the clamped 120% high neutron flux

scram setpoint which is the only APRM scram setpoint with any

analytical safety basis, the proposed changes will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously evaluated

because the proposed changes do not allow plant operation in any

mode that is not already evaluated. The APRM system provides

monitoring and accident mitigation functions to limit peak flux in

the core during Modes 1 and 2. No pressure boundary interfaces or

process control parameters will be challenged in any way as to

create the possibility of a new or different type of accident than

any previously evaluated. Also, failure of the sensing line

associated with flow transmitters to measure recirculation drive

flow has already been accounted for in the initial plant design by

including excess flow check valves for sensing line break isolation.

Therefore, these changes will not create the possibility of a new or

different kind of accident than any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety because the APRM flow biased high flux scram

is not credited in the PBAPS safety analysis. Because the proposed

changes do not impact safety analysis assumptions, these 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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105.

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101

NRC Project Director: John F. Stolz

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: April 22, 1997

Description of amendment request: The proposed amendment would

revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)

Section 4.2.b, ``Steam Generator Tubes,'' to allow a laser-welded

repair of Westinghouse hybrid expansion joint (HEJ) sleeved steam

generator (SG) tubes. The proposed repair process would fuse the tube

to the sleeve in the upper joint of the existing HEJ sleeved tubes. The

repair weld would be made in either the hardroll (HR) expansion or the

upper hydraulic expansion (HE) region of the HEJ. By fusing the tube to

the sleeve, parent tube degradation below the weld would be isolated

and a new pressure boundary would be formed. The new pressure boundary

would satisfy both the structural and leakage integrity requirements of

the sleeved tube assembly with no change in the flow or heat transfer

characteristics of the sleeved tube. The proposed amendment supersedes

in its entirety a previously submitted proposed amendment dated

September 6, 1996, which was noticed in the Federal Register on October

15, 1996 (61 FR 53769).

[[Page 24989]]

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. Operation of the KNPP in accordance with the proposed license

amendment does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The laser-weld repair of HEJ sleeved tubes in either the HR or

HE location will not affect the tube, sleeve, or weld stress

conditions or fatigue usage factors such that the limits of the ASME

Boiler and Pressure Vessel Code are exceeded. Accelerated corrosion

testing performed on prototypic HR welds, and a corrosion assessment

performed for the HE welds concluded that the repair welds will not

result in aggravated stress corrosion cracking at the weld-repair

location. Any postulated sleeve joint degradation would occur at a

relatively slow rate and would be detectable by routine non-

destructive examination (NDE) inspection prior to reaching any

applicable safety margins. Therefore, use of the laser-weld repair

process will not result in an increased probability of an accident

previously evaluated.

A post-weld stress relief ultrasonic test inspection is required

to verify minimum acceptable weld thickness to ensure that the weld

stresses do not exceed ASME Code limits for both stress intensity

and fatigue usage. Leakage testing of laser-welded sleeve joints,

and in-situ leakage testing of the laser-welded repairs (LWR) at

KNPP, demonstrate a leak-tight joint at pressures up to main steam

line break. Mechanical testing of 7/8 inch laser-welded tubesheet

sleeves installed in roll-expanded tubes has shown that the

individual joint structural strength of Alloy 690 laser-welded

sleeves under normal, upset, and faulted conditions provides margin

to acceptable limits. These acceptable limits bound the most

limiting (3 times normal operating pressure differential)

recommended by Regulatory Guide (RG) 1.121.

The HEJ sleeve plugging limit currently defined in the TS is

reduced from 31% to 24% throughwall due to the use of ASME code

minimum material properties values for the sleeve material. Minimum

wall thickness requirements (used for developing the depth-based

plugging limit for the sleeve) are determined using the guidance of

RG 1.121 and the pressure stress equation of Section 3 of the ASME

Code.

The hypothetical consequences of failure of the laser-welded

repaired HEJ would be bounded by the current SG tube rupture (SGTR)

analysis covered in the KNPP Updated Safety Analysis Report. Due to

the slight reduction in diameter caused by the sleeve wall

thickness, primary coolant release rates would be slightly less than

assumed for the SGTR, and, therefore, would result in lower primary

fluid mass release to the secondary system. The laser-weld repair

process does not change the existing reactor coolant system flow

conditions; therefore, existing loss of coolant accident (LOCA) and

non-LOCA analysis results will be unaffected. Plant response to

design basis accidents for the current tube plugging and flow

conditions are not affected by the repair process; no new tube

diameter restrictions are introduced. Therefore, the application of

the repair weld will not increase the consequences of a previously

evaluated accident.

2. The proposed license amendment request does not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

Application of laser-welded repair for the HEJ sleeved tubes

will not introduce significant or adverse changes to the plant

design basis. The general configuration of the HEJ sleeve is

unaffected by the repair process. The repair process also does not

represent a potential to affect any other plant component. Stress

and fatigue analysis of the repair has shown that the ASME Code and

RG 1.121 criteria are not exceeded. Application of the laser-weld

repair to the HEJ sleeved tubes maintains overall tube bundle

structural and leakage integrity. Extensive testing and evaluation

including examination of actual pulled tube samples verified

adequate structural and leakage integrity of repair HEJs, which had

acceptable NDE.

Any hypothetical accident as a result of potential tube or

sleeve degradation in the repaired portion of the joint is bounded

by the existing tube rupture accident analysis. Therefore, use of

the laser-welded repair process will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed license amendment does not involve a significant

reduction in the margin of safety.

The laser-weld repair of the HEJ sleeved tubes has been shown to

restore integrity of the tube bundle consistent with its original

design basis conditions; i.e., tube/sleeve operational and faulted

load stresses and cumulative fatigue usage factors are bounded by

ASME Code requirements and the tubes are leak tight under all plant

conditions. Based on the results of the structural and leakage

testing performed on LWR joints pulled from the KNPP SGs and

supporting analytical evaluations, application of laser-welded

repair will 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.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P.O. Box 1497, Madison, Wisconsin 53701-1497.

NRC Project Director: Gail H. Marcus

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: April 24, 1997

Description of amendment request: The proposed amendment would

revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)

Section 4.2.b, ``Steam Generator Tubes,'' to allow repair of steam

generator (SG) tubes with Combustion Engineering (CE) leak-tight

sleeves in accordance with CE generic topical report CEN-629-P,

Revision 2, ``Repair of Westinghouse Series 44 and 51 Steam Generator

Tubes Using Leak-Tight Sleeves.'' The TS would also be revised to allow

re-sleeving of tubes with existing sleeve joints in accordance with

KNPP specific topical report CEN-632-P, ``Repair of Kewaunee Steam

Generator Tubes Using a Re-Sleeving Technique.''

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. Operation of the KNPP in accordance with the proposed license

amendment does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The supporting technical evaluation and safety evaluation for

the CE leak-tight sleeves demonstrates that the sleeve configuration

will provide SG tube structural and leakage integrity under normal

operating and accident conditions. The sleeve configurations have

been designed and analyzed in accordance with the requirements of

the ASME Code. Mechanical testing has shown that the sleeve and

sleeve joints provide margin above acceptance limits. Ultrasonic

testing is used to verify the leak tightness of the weld above the

tubesheet. Testing has demonstrated the leak tightness of the

hardroll joint as well as the structural integrity of the hardroll

joint. Tube rupture cannot occur at the hardroll joint due to the

reinforcing effect of the tubesheet. Tests have demonstrated that

tube collapse will not occur due to postulated loss of coolant

accident loadings.

The existing TS leak-rate requirements and accident analysis

assumptions remain unchanged in the event that significant leakage

does occur from the sleeve joint or the sleeve assembly ruptures.

Any leakage through the sleeve assembly is fully bounded by the

existing SG tube rupture analysis included in the KNPP Updated Final

Safety Analysis Report. The proposed sleeving and re-sleeve repair

processes do not adversely impact any other previously evaluated

design basis accidents.

2. The proposed license amendment request does not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

[[Page 24990]]

Installation of the sleeves or re-sleeves does not introduce any

significant changes to the plant design basis. The use of a sleeve

to span the area of degradation of the SG tube restores the

structural and leakage integrity of the tubing to meet the original

design basis. Stress and fatigue analysis of the sleeve assembly

shows that the requirements of the ASME Code are met. Mechanical

testing has demonstrated that margin exists above the design

criteria. Any hypothetical accident as a result of any degradation

in the sleeved tube would be bounded by the existing tube rupture

accident analysis.

3. The proposed license amendment does not involve a significant

reduction in the margin of safety.

The use of sleeves to repair degraded SG tubing has been

demonstrated to maintain the integrity of the tube bundle

commensurate with the requirements of the ASME Code and draft

Regulatory Guide 1.121, and to maintain the primary to secondary

pressure boundary under normal and postulated accident conditions.

The safety factors used in the verification of the strength of the

sleeve assembly are consistent with the safety factors in the ASME

Boiler and Pressure Vessel Code used in SG design. The operational

and faulted condition stresses and cumulative usage factors are

bounded by the ASME Code requirements. The sleeve assembly has been

verified by testing to prevent both tube pullout and significant

leakage during normal and postulated accident conditions. A test

program was conducted to ensure the lower hardrolled joint design

was leak tight and capable of withstanding the design loads. The

primary coolant pressure boundary of the sleeve assembly will be

periodically inspected by non-destructive examination to identify

sleeve degradation due to operation.

Installation of the sleeves and re-sleeves will decrease the

number of tubes that must be taken out-of-service due to plugging.

There is a small amount of primary coolant flow reduction due to the

sleeve for which an equivalent plugging sleeve to plug ratio is

assigned based on sleeve length. The ratio is used to assess the

final equivalent plugging percentage as an input to other safety

analyses. Because the sleeve maintains the design basis requirements

for the SG tubing, it is concluded that 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.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P.O. Box 1497, Madison, Wisconsin 53701-1497.

NRC Project Director: Gail H. Marcus

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.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Elecric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendments request: March 27, 1997

Brief description of amendments: The proposed amendments would

revise the Technical Specifications for the Brunswick Steam Electric

Plant Units 1 and 2 to eliminate certain instrumentation response time

testing requirements in accordance with NRC-approved BWR Owners Group

Topical Report NEDO-32291-A, ``System Analysis for the Elimination of

Selected Response Time Testing Requirements.''

Date of publication of individual notice in Federal Register: April

1, 1997(62 FR 15542)

Expiration date of individual notice: May 1, 1997

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Northern States Power Company, Docket Nos. 50-282 and 50-306,

Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue

County, Minnesota, and Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of amendment requests: December 6, 1996

Description of amendment requests: The licensee requests amendments

to the Prairie Island and Monticello operating licenses to reflect the

Commission's approval of the transfer of control over the subject NRC

licenses held by Northern States Power Company (NSP). On October 20,

1995, as supplemented August 28, 1996, NSP requested NRC approval for

the transfer of control of licenses. The Commission is considering the

issuance of amendments to the licenses to reflect the above transfer

approved by the Commission on April 1, 1997 (62 FR 17882, dated April

11, 1997).

Date of individual notice in the Federal Register: April 11, 1997

(62 FR 17882)

Expiration date of individual notice: May 12, 1997

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: April 4, 1997

Brief description of amendment request: The proposed amendment

would clarify the scope of the surveillance requirements for response

time testing of instrumentation in the reactor protection system,

isolation actuation system, and emergency core cooling system in the

Technical Specifications for each unit (Sections 4.3.1.3, 4.3.2.3, and

4.3.3.3).

Date of publication of individual notice in Federal Register: April

17, 1997 (62 FR 17885)

Expiration date of individual notice: May 19, 1997

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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

[[Page 24991]]

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 at the local public document rooms for

the particular facilities involved. Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of application for amendment: January 24, 1997, as

supplemented March 27, 1997

Brief description of amendment: The proposed amendment will update

the Safety Limit Minimum Critical Power Ratio (SLMCPR) in Technical

Specification 2.1.2 and the associated Bases section to reflect the

results of the latest cycle-specific calculation performed for the

Pilgrim Nuclear Power Station Operating Cycle 12. In addition, the

values provided in Note 5 of Table 3.2.C.1, which are based on the

SLMCPR values, have been revised as a result of the changes to the

SLMCPR value.

Date of issuance: April 7, 1997

Effective date: April 7, 1997

Amendment No.: 171

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 12, 1997 (62

FR 6568) The March 27, 1997, supplemental letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated April 7, 1997 No

significant hazards consideration comments received: No

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of application for amendment: March 27, 1997, as supplemented

April 11, 1997.

Brief description of amendment: The amendments revise the Technical

Specifications relating to response time testing requirements

associated with the reactor protection system, isolation system, and

emergency core cooling system.

Date of issuance: April 18, 1997

Effective date: April 18, 1997

Amendment Nos.: 184 and 215

Facility Operating License Nos. DPR-71 and DPR-62. Amendments

revised the Technical Specifications. Public comments requested as to

proposed no significant hazards consideration (NSHC): Yes (62 FR 15542

dated April 1, 1997). The notice provided an opportunity to submit

comments on the Commission's proposed NSHC determination. No comments

have been received. The notice also provided for an opportunity to

request a hearing by May 1, 1997, but indicated that if the Commission

makes a final NSHC determination, any such hearing would take place

after issuance of the amendments. The Commission's related evaluation

of the amendment, finding of exigent circumstances, and final

determination of NSHC are contained in a Safety Evaluation dated April

18, 1997.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of application for amendments: December 21, 1995, as

supplemented on October 24, 1996, and March 24, 1997.

Brief description of amendments: The amendments relocate certain

cycle-specific parameter limits from the Technical Specifications (TS)

to the Operating Limits Report. The cycle-specific parameter limits to

be relocated are for Shutdown Rod Insertion Limit, Control Rod

Insertion Limits, Axial Flux Difference Target Band, Heat Flux Hot

Channel Factor [FQ(z)], and Nuclear Enthalpy Rise Hot

Channel Factor (FN delta H). In addition, your March 24,

1997, submittal contained supplementary revisions to the Bases section

associated with the above TS change. The supplementary Bases pages will

be reviewed and transmitted to you under separate cover. Finally,

Braidwood's TS 6.9.1.7 title was corrected.

Date of issuance: April 16, 1997

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 88, 88, 80, 80

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 20, 1997 (62

FR 7804). The March 24, 1997, submittal 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 April 16, 1997. No

significant hazards consideration comments received: No

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of application for amendments: April 29, 1996, as supplemented

on January 21 and March 25, 1997.

Brief description of amendments: The amendments would: (1) revise

Technical Specification (TS) 3.7.1.1, Action a., to require the unit to

be in hot shutdown, rather than cold shutdown, for consistency with

NUREG-1431, ``Standard Technical Specifications for Westinghouse

Plants,'' and add a new Action b. to clarify the shutdown requirements

when there are more than three inoperable main steam line American

Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code

(Code) safety valves on any

[[Page 24992]]

one steam generator; (2) revise TS Surveillance Requirement 4.7.1.1 to

clarify that Specification 4.0.4 does not apply for entry into Mode 3

for Byron and Braidwood and for Braidwood only, delete the one-time

requirements for Unit 1, Cycle 5 and Unit 2 after outage A2F27; (3)

revise the maximum allowable power range neutron flux high trip

setpoints in Table 3.7-1; (4) revise Table 3.7-2 to increase the as-

found main steam safety valve (MSSV) lift setpoint tolerance to plus or

minus 3 percent, provide an as-left setpoint tolerance of plus or minus

1 percent, and change a table notation; (5) delete the orifice size

column from Table 3.7-2; and (6) revise the Bases for TS 3.7.1.1 to be

consistent with the proposed changes to TS 3.7.1.1.

Date of issuance: April 15, 1997

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 87, 87, 79, and 79

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 12, 1997 (62 FR

11486). The March 25, 1997, submittal provided additional 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 April 15, 1997. No

significant hazards consideration comments received: No

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: August 7, 1996, as supplemented

March 12, 1997.

Brief description of amendment: The amendment revises Technical

Specifications to allow the use of 10 CFR Part 50, Appendix J, Option

B, ``Performance-Based Containment Leak Rate Testing.''

Date of issuance: April 10, 1997

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 190

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 11, 1996 (61

FR 47976) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 10, 1997. No significant

hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,

Michigan

Date of application for amendment: March 27, 1997, as supplemented

on April 4, 1997

Brief description of amendment: The amendment revises technical

specification surveillance requirement (SR) 4.3.1.3 for the Reactor

Protection System Instrumentation to indicate that certain sensors are

exempt from response time testing. A similar revision is made to SR

4.3.2.3 for the Isolation Actuation Instrumentation. Finally, SR

4.3.3.3 for the Emergency Core Cooling System Actuation Instrumentation

is revised to indicate that the emergency core cooling system actuation

instrumentation is exempt from response time testing.

Date of issuance: April 18, 1997

Effective date: April 18, 1997, with full implementation prior to

entry into Operation Condition 2 or 3

Amendment No.: 111

Facility Operating License No. NPF-43. Amendment revises the

Technical Specifications.

Public comments requested as to proposed no significant hazards

considerations (NSHC): Yes (62 FR 15731 dated April 2, 1997). The

notice provided an opportunity to submit comments on the Commission's

proposed NSHC determination. No comments have been received. The notice

also provided for an opportunity to request a hearing by May 2, 1997,

but indicated that if the Commission makes a final NSHC determination,

any such hearing would take place after issuance of the amendment. The

Commission's related evaluation of the amendment, finding of exigent

circumstances, and final determination of NSHC are contained in a

Safety Evaluation dated April 18, 1997.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina Date of

application for amendments: January 6, 1997, as supplemented by

letters dated April 10 and 15, 1997

Brief description of amendments: The amendments revise portions of

the Technical Specifications to permit a one-time operation of the

Containment Purge Ventilation System during Modes 3 and 4 after the

current and forthcoming steam generator replacement outages.

Date of issuance: April 24, 1997

Effective date: As of the date of issuance to be implemented within

30 days

Amendment Nos.: 174 and 156

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 12, 1997 (62

FR 6574) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 24, 1997. No significant

hazards consideration comments received: No

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina 28223-0001

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of application for amendments: September 9, 1996

Brief description of amendments: These amendments modify the design

features section (Section 5.0) of the Technical Specifications (TSs) to

make the design features section consistent with the intent of 10 CFR

50.36 and with the guidance provided in the NRC's Standard Technical

Specifications, Westinghouse Plants (NUREG-1431, Revision 1).

Date of issuance: April 14, 1997

Effective date: Both units, as of date of issuance, to be

implemented within 60 days.

Amendment Nos.: 202 and 83

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64384) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 14, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: B. F. Jones Memorial Library,

[[Page 24993]]

663 Franklin Avenue, Aliquippa, PA 15001

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,

Unit No. 2, Pope County, Arkansas

Date of application for amendment: December 19, 1996

Brief description of amendment: The amendment deletes the specific

value for the total reactor coolant system volume from the Design

Features section of the Technical Specifications.

Date of issuance: April 16, 1997

Effective date: April 16, 1997

Amendment No.: 181

Facility Operating License No. NPF-6: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4348) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 16, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,

Unit No. 2, Pope County, Arkansas

Date of application for amendment: December 19, 1996

Brief description of amendment: Request to add CENTS code as a

Reference to the Technical Manual used for determining Core Operating

Limits Report in the Technical Specifications.

Date of issuance: April 24, 1997

Effective date: April 24, 1997

Amendment No.: 182

Facility Operating License No. NPF-6: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4347) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 24, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station,

Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: January 10, 1997

Brief description of amendment: The amendment revises the technical

specifications for reactor pressure vessel pressure and temperature

limits by providing new limits that are valid to 12 effective full

power years.

Date of issuance: April 14, 1997

Effective date: April 14, 1997

Amendment No.: 93

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 26, 1997 (62

FR 8798) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 14, 1997. No significant

hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: November 7, 1995, as supplemented by

letters dated July 17, and December 26, 1996, and February 27, March

14, April 7, and April 17, 1997.

Brief description of amendment: The amendment changes the Appendix

A Technical Specifications by revising TS 3/4.8.1, ``Electrical Power

Systems - A.C. Sources,'' to incorporate recommendations and

suggestions from (1) Generic Letter (GL) 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements for

Testing During Power Operations;'' (2) GL 94-01, ``Removal of

Accelerated Testing and Special Reporting Requirements for Emergency

Diesel Generators from Plant Technical Specifications;'' and (3) NUREG-

1432, ``Standard Technical Specifications Combustion Engineering

Plants.''

Date of issuance: April 21, 1997

Effective date: April 21, 1997, to be implemented within 60-days.

Amendment No.: 126

Facility Operating License No. NPF-38: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 3, 1996 (61 FR

180) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 21, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: October 10, 1996 (TSCR 243)

Brief description of amendment: The amendment modifies the

Technical Specifications (TS) by replacing the description of the

existing permissive interlock from AC Voltage to Core Spray Booster

Pump d/p Permissive: 21.2 psid for initiation of the

automatic depressurization system, adds corresponding surveillance

requirements, and adds notes clarifying functional requirements.

Date of Issuance: April 14, 1997

Effective date: April 14, 1997, with full implementation within 60

days

Amendment No.: 190

Facility Operating License No. DPR-16.

Date of initial notice in Federal Register: November 6, 1996 (61 FR

57485). The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated April 14, 1997 No significant

hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New London County, Connecticut

Date of application for amendment: September 5, 1996

Brief description of amendment: The amendment deletes License

Condition 2.C.(5), ``Integrated Implementation Schedule'' from the

Millstone Unit 1 Operating License.

Date of issuance: April 15, 1997

Effective date: As of the date of issuance, to be implemented

within 60 days.

Amendment No.: 100

Facility Operating License No. DPR-21: Amendment revised the

Operating License.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55036) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 15, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut 06360 and at the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Northeast Nuclear Energy Company, et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of application for amendment: February 5, 1996

[[Page 24994]]

Brief description of amendment: The amendment deletes a clause from

Technical Specification 4.0.5.a. Specifically, this change deletes the

clause ``(g), except where specific written relief has been granted by

the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).''

The amendment also makes the appropriate changes to the Bases section.

Date of issuance: April 21, 1997

Effective date: As of the date of issuance, to be implemented

within 60 days.

Amendment No.: 138

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 26, 1997 (62

FR 8800) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 21, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Northeast Nuclear Energy Company, et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of application for amendment: March 4, 1996

Brief description of amendment: The amendment modifies Surveillance

Requirements 4.8.1.1.2.a.6, 4.8.1.1.2.b, and 4.8.1.1.2.g.7 by

specifying load bands in loading the diesel generator (DG) in lieu of

the present requirement to load the DG greater than or equal to a given

value. A footnote is being added to the three surveillance

rerquirements to indicate that a momentary transient outside the load

range shall not invalidate the test. The aassociated Bases sections

have been revised to reflect the above changes.

Date of issuance: April 15, 1997

Effective date: As of the date of issuance, to be implemented

within 60 days.

Amendment No.: 137

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 12, 1997 (62 FR

11496) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 15, 1997. No significant

hazards consideration comments received: No

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,

Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County, California

Date of application for amendments: February 14, 1996, as

supplemented by letter dated February 24, 1997.

Brief description of amendments: The amendments revised the

combined Technical Specifications (TS) for the Diablo Canyon Power

Plant (DCPP) Unit Nos. 1 and 2 to revise 30 TS and add two new TS

surveillance requirements to support implementation of extended fuel

cycles at DCPP Unit Nos. 1 and 2. The specific TS changes include those

for 9 trip actuating device tests, 12 fluid system actuation tests, and

11 miscellaneous tests. Two of the fluid system actuation tests are new

TS surveillance requirements. The TS changes also involve adding a new

frequency notation, ``R24, REFUELING INTERVAL,'' to Table 1.1 of the

TS. Also, a revision that applies to all subsequent TS changes involves

revising the Bases Section of TS 4.0.2 to change the surveillance

frequency from an 18-month surveillance interval to at least once each

refueling interval.

Date of issuance: April 14, 1997

Effective date: April 14, 1997, to be implemented within 90 days

from the date of issuance.

Amendment Nos.: Unit 1 - 118; Unit 2 - 116

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 19, 1996 (61 FR

31183) The February 24, 1997, supplemental letter provided additional

clarifying information and did not change the staff's initial no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated April 14, 1997. No significant hazards consideration

comments received: No.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,

Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County, California

Date of application for amendments: May 31, 1996, as supplemented

by letter dated December 16, 1996.

Brief description of amendments: The amendments revised the

combined Technical Specifications (TS) for the Diablo Canyon Power

Plant (DCPP) Unit Nos. 1 and 2 to revise 23 TS surveillance frequencies

from at least once every 18 months to at least once per refueling

outage (nominally 24 months) and to make administrative changes for 6

other TS to maintain consistency for TS that are not proposed for

surveillance extension. The specific TS changes proposed include those

for 2 response time tests, 3 containment spray system tests, and 24

ventilation system tests.

Date of issuance: April 14, 1997

Effective date: April 14, 1997, to be implemented within 90 days of

issuance.

Amendment Nos.: Unit 1 - 119; Unit 2 - Amendment No. 117

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 9, 1996 (61 FR

52966) The December 16, 1996, supplemental letter provided additional

clarifying information and did not change the staffs initial no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated April 14, 1997. No significant hazards consideration

comments received: No.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: November 22, 1996

Brief description of amendment: The amendment allows an increase in

the U-235 enrichment of fuel stored in the fresh fuel storage racks or

the spent fuel storage racks from 4.5 weight percent (w/o) U-235 to 5.0

w/o U-235.

Date of issuance: April 15, 1997

Effective date: As of the date of issuance to be implemented within

30 days.

[[Page 24995]]

Amendment No.: 173

Facility Operating License No. DPR-64: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 15, 1997 (62 FR

2182) The Commission's related evaluation of the amendment is contained

in the Safety Evaluation dated April 15, 1997, and an Environmental

Assessment dated March 25, 1997. No significant hazards consideration

comments received: Yes

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch

Nuclear Plant, Unit 2, Appling County, Georgia

Date of application for amendments: December 3, 1996, as

supplemented by letters dated January 27 and April 4, 1997

Brief description of amendments: The amendments revise Technical

Specification 2.1.1.2 to change the Safety Limit Minimum Critical Power

Ratio based on the cycle-specific analyses of Cycle 13 of a non-

equilibrium core of all General Electric (GE) 9 fuel with varying

enrichments and Cycle 14 of a non-equilibrium mixed core of GE13 and

GE9 fuel.

Date of issuance: April 17, 1997

Effective date: For Cycle 13, as of the date of issuance; For Cycle

14, effective upon startup.

Amendment Nos.: 148 for Cycle 13; 149 for Cycle 14

Facility Operating License Nos. DPR-57 and NPF-5. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4349) The January 27 and April 4, 1997, letters provided additional

information that did not change the scope of the December 3, 1996,

application and the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated April 17, 1997. No

significant hazards consideration comments received: No

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: April 4, 1996, as supplemented

by letters dated January 10, February 7, February 13, March 17, March

19, March 20, March 25, April 1, April 6, April 10, April 11, and April

18, 1997.

Brief description of amendments: The amendments revise the Sequoyah

Technical Specifications (TSs) and associated Bases to allow for the

conversion from Westinghouse fuel to Framatome Cogema Fuel, designated

Mark-BW. The planned fuel conversion begin with fuel cycle 9 for each

unit. The amendments would revise the TSs to reflect the fuel design

and vendor change. The licensee's evaluation was contained in Topical

Report BAW-10220P, ``Mark-BW Fuel Assembly Application for Sequoyah

Nuclear Units 1 and 2.''

Date of issuance: April 21, 1997

Effective date: As of the date of issuance to be implemented no

later than 45 days of its issuance for Unit 1, and implemented upon

installation of Framatome Cogema Fuel in the Unit 2 reactor vessel for

Unit 2.

Amendment Nos.: 223 and 214

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the Technical Specifications and License Conditions.

Date of initial notice in Federal Register: May 8, 1996 (61 FR

20856) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 21, 1997 No significant

hazards consideration comments received: No

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

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 for the

amendment 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.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the

[[Page 24996]]

documents related to this action. Accordingly, the amendments have been

issued and made effective 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

application for amendment, (2) the amendment to Facility Operating

License, 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 at

the local public document room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By June 6, 1997, 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 at the local public document room for the

particular facility involved. 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 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. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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 20555-001, 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. Where petitions are filed during the last 10 days of the notice

period, it is requested that the petitioner promptly so inform the

Commission by a toll-free telephone call to Western Union at 1-(800)

248-5100 (in Missouri 1-(800) 342 6700). The Western Union operator

should be given Datagram Identification Number N1023 and the following

message addressed to (Project Director): petitioner's name and

telephone number, date petition was mailed, plant name, and publication

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-001, 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 the

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Unit Nos. 2 and 3, Grundy County, Illinois

Date of application for amendments: April 14, 1997, as supplemented

on April 17, April 22, and April 24, 1997.

Brief description of amendments: The proposed amendments requested

(1) review and approval of an Unreviewed Safety Question (USQ)

involving the control room operator dose resulting from an error in the

secondary containment volume, (2) a change in Technical Specification

(TS) Surveillance Requirements (SR) 4.7. P.2.b and 4.7. P.3 values for

the allowed methyl iodide penetration for the standby gas treatment

charcoal adsorbers, and (3) change of TS 5.2.C to

[[Page 24997]]

reflect the new calculated free volume of the secondary containment.

The April 17, April 22 and April 24, 1997, submittals provided

additional clarifying information that did not change the initial

proposed no significant hazards consideration determination.

Date of Issuance: April 25, 1997

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 158 and 153

Facility Operating License Nos. DPR-19 and DPR-25: The amendments

revised the Technical Specifications. Press release issued requesting

comments as to proposed no significant hazards consideration: Yes.

April 22, 1997. Joliet Herald News. Comments received: No. The

Commission's related evaluation of the amendments, finding of exigent

circumstances, consultation with the State of Illinois and final

determination of no significant hazards consideration are contained in

a Safety Evaluation dated April 25, 1997.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60690

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450

NRC Project Director: Robert A. Capra

Pennsylvania Power and Light Company, Docket No. 50-388,

Susquehanna Steam Electric Station, Unit 2, Luzerne County,

Pennsylvania

Date of application for amendment: April 16, 1997, and as

supplemented by a letter dated April 18, 1997

Brief description of amendment: This amendment changes the footnote

in the Design Features Section 5.3.1 of the Technical Specifications to

allow the use of ATRIUM-10 fuel in Operational Conditions 3 and 4.

Date of issuance: April 25, 1997

Effective date: As of the date of issuance to be implemented upon

receipt.

Amendment No.: 138

Facility Operating License No. NPF-22: This amendment revised the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes. The NRC published a public

notice of the proposed amendment, issued a proposed finding of no

significant hazards consideration and reqeusted that any comments on

the proposed no significant hazards consideration be provided to the

staff by the close of business on April 24, 1997. The notice was

published in the Wilkes-Barre Times Leader and the Berwick Press

Enterprise on April 22-24, 1997. Public comments were received and have

been addressed in the staff's safety evaluation.

The Commission's related evaluation of the amendment, finding of

exigent circumstances, consultation with the State of Pennsylvania and

final no significant hazards consideration determination are contained

in a Safety Evaluation dated April 25, 1997.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

and Trowbridge, 2300 N Street NW., Washington, DC 20037

NRC Project Director: John F. Stolz

Dated at Rockville, Maryland, this 30th day of April 1997.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects III/IV, Office of Nuclear

Reactor Regulation.

[Doc. 97-11725 Filed 5-6-97; 8:45 am]

BILLING CODE 7590-01-F