Biweekly Notice; 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 Pub. L. 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

[[Page 40904]]

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 July 3, 1999, through July 16, 1999. The

last biweekly notice was published on July 14, 1999 (64 FR 38022).

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 and

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and should cite the publication date and page number of

this Federal Register notice. Written comments may also be delivered to

Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of

written comments received may be examined at the NRC Public Document

Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The

filing of requests for a hearing and petitions for leave to intervene

is discussed below.

By August 27, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating 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 much 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.

[[Page 40905]]

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: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Stream

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: March 26, 1999.

Description of amendment request: The proposed change provides a

Required Action and Completion Time for the Ultimate Heat Sink (UHS) in

the event that service water temperature exceeds the current 95 deg.F

surveillance limit. It involves an allowance to continue operation for

a period of 8 hours with the UHS at a temperature greater than the

temperature limits provided in Technical Specification (TS) Limiting

Condition of Operation 3.7.8, ``Ultimate Heat Sink (UHS)'' and provides

an upper UHS temperature limit beyond which plant shutdown is required.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Carolina Power & Light (CP&L) Company has evaluated the proposed

Technical Specification change and has concluded that it does not

involve a significant hazards consideration. The conclusion is in

accordance with the criteria set forth in 10 CFR 50.92. The bases

for the conclusion that the proposed change does not involve a

significant hazards consideration are discussed below.

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures or components. The proposed change will

allow plant operation for a short period of time when the service

water temperature exceeds 95 deg.F. If the service water temperature

is restored within the allowed time, a plant shutdown is not

required. This minimizes plant transients, which reduces the

probability of a reactor trip and the resulting challenges to

mitigating systems. A service water temperature of up to 99 deg.F

does not increase the failure rate of systems, structures or

components because the systems, structures, and components are

designed for higher temperatures than at which they operate.

The Service Water (SW) System temperature is not assumed to be

an initiating condition of any accident evaluated in the safety

analysis report. Therefore, the allowance of a limited time for

service water temperature to be in excess of 95 deg.F does not

involve an increase in the probability of an accident previously

evaluated in the safety analysis report (SAR). The SW System

supports operability of safety related systems used to mitigate the

consequences of an accident. The service water temperature is not

expected to increase significantly beyond 95 deg.F due to the

limited time allowed by the proposed change in conjunction with the

generally slow rate of temperature increase experienced from thermal

changes in Lake Robinson. The capability of components to perform

their safety related function is not affected up to a service water

temperature of 99 deg.F with the exception of the Containment Air

Recirculation Fan Coolers. The heat removal capacity of the

Containment Air Recirculation Fan Coolers is not expected to be

significantly reduced by a small increase in service water

temperature. If heat removal is not significantly reduced,

containment pressure and leakage will not be significantly

increased, and the doses from containment leakage will not be

significantly increased. Therefore, the proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated in the SAR.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures or components. A service water temperature

of up to 99 deg.F does not introduce new failure mechanisms of

systems, structures or components not already considered in the SAR

because the systems, structures, and components are designed for

higher temperatures than at which they operate. Therefore, the

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

previously evaluated is not created.

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

of safety?

The proposed change will allow a small increase in service water

temperature above the design basis limit for the SW System and delay

by 8 hours the requirement to shutdown the plant when the service

water system design limit is exceeded. There are design margins

associated with systems, structures and components that are cooled

by the service water system that are affected. The capability of

components to perform their safety related function is not affected

up to a service water temperature 99 deg.F with the exception of the

Containment Air Recirculation Fan Coolers. The Containment Air

Recirculation Fan Coolers remove heat from containment to mitigate

containment pressure and temperature following a MSLB (main

streamline break) inside containment or a Large Break LOCA (loss-of-

coolant accident) inside containment. An increase in service water

temperature in excess of the design limit due to hot weather

conditions is expected to be small due to the limited time allowed

by the proposed change in conjunction with the generally slow rate

of temperature increase experienced from thermal changes in Lake

Robinson. Therefore, the effect on the Containment Air Recirculation

Fan Coolers' heat removal capacity and the resulting containment

pressure and temperature is expected to be small. Therefore, there

is no significant reduction in margin of safety associated with this

change.

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: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602.

NRC Section Chief: Sheri R. Peterson.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: June 29, 1999.

Description of amendment request: This amendment request proposes

to increase the notch testing surveillance interval of partially

withdrawn control rods in Technical Specification Surveillance

Requirement 3/4.3.C,

[[Page 40906]]

``Reactivity Control--Control Rod Operability,'' from an interval of

once in 7 days to once in 31 days.

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:

Does the change involve a significant increase in the

probability of occurrence or consequences of an accident previously

evaluated?

The proposed change extends the Surveillance Frequency for

partially withdrawn control rods. The change does not affect

equipment design or operation. The affected Surveillance is not

considered to be an accident initiator. Therefore, this change will

not significantly increase the probability of an accident previously

evaluated. Furthermore, extension of the Surveillance Frequency will

not impact the ability to perform its function following an

accident.

Therefore, the proposed changes do not create the possibility of

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

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The extension of the Surveillance Frequency does not involve

physical modification to the plant and does not introduce a new mode

of operation.

Therefore, the change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

Does the change involve a significant reduction in a margin of

safety?

The change in the Surveillance Frequency only provides a minor

reduction in the probability of finding an inoperable control rod.

Most of the control rods will continue to be tested on the current

Frequency. However, if one stuck rod is identified, all rods must be

checked promptly.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff proposed to determine that

the requested amendments involve no significant hazards

consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear

Station, Unit 1, Dauphin County, Pennsylvania

Date of amendment request: April 1, 1999.

Description of amendment request: The proposed license amendment

would modify the Technical Specifications (TSs) to incorporate certain

improvements from the Revised Standard Technical Specifications for B&W

Plants (NUREG-1430) that would add limiting conditions for operation

action statements, make surveillance requirements more consistent with

the revised standard TSs, correct conflicts or inconsistencies from

earlier TS revisions, correct administrative errors, and revise the

spent fuel pool sampling from monthly and after adding chemicals to

weekly.

The staff's proposed no significant hazards determination below

does not address the licensee's proposed changes with respect to a high

pressure injection system operation in a low temperature overpressure

environment.

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 facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or consequences of an accident previously

evaluated. The proposed amendment makes administrative corrections,

adds conditions to the limiting conditions of operation [LCOs],

revises selected time clocks and surveillance requirements

consistent with NUREG 1430, and adds a time clock to a unique LCO.

These changes have no effect on the plant design or operation. The

reliability of systems and components relied upon to prevent or

mitigate the consequences of accidents previously evaluated is not

degraded by proposed changes. Therefore, operation in accordance

with the proposed amendment does not involve a significant increase

in the probability of occurrence or consequences of an accident

previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any previously evaluated, because no new

accident initiators would be created.

3. Operation of the facility in accordance with the proposed

amendment will not involve a significant reduction in a margin of

safety because no changes to plant operating limits or limiting

safety system settings are proposed.

The NRC staff has reviewed the licensee's analysis and based on the

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: Law/Government Publications

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

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,

Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 22037.

NRC Section Chief: S. Singh Bajwa.

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 amendment request: June 4, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 Technical Specifications (TSS) proposes to revise

the definition of operating personnel in section 6.2.2.g to make it

consistent with the Standard Technical Specifications and to remove a

footnote.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licenses has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

No, these TS changes are administrative in nature. Removing the

statement in section 6.2.2.g that defines on shift operating

personnel and adding a new paragraph consistent with the Standard

Technical Specifications is an administrative line item change that

follows NRC guidance. The current statement is not needed because TS

Table 6.2.1 defines the minimum operations shift crew composition

and commitments to Table B-1 of NUREG-0654 defines the minimum

staffing requirements for each function area.

The change to TS 6.2.2.i is administrative in nature. The

statement that reads, ``For the period ending three years after

restart from the 1993/1994 Performance Improvement Outage, the

Operations Manager will be permitted to have held a SRO [senior

reactor operator] license at a Pressurized Water Reactor other than

Indian Point Unit 3'', was a relaxation of the requirements of

6.2.2i.

Therefore, these changes will not increase the probability or

consequences of an accident previously evaluated, because they are

administrative and affect neither accident initiation or mitigation.

2. Does the proposed license amendment create the possibility of

a new or different

[[Page 40907]]

kind of accident from any accident previously evaluated?

No, these TS changes are administrative in nature. Removing the

statement in section 6.2.2.g that defines on shift operating

personnel and adding a new paragraph consistent with the Standard

Technical Specifications is an administrative line item change that

follows NRC guidance. The current statement is not needed because TS

Table 6.2-1 defines the minimum operations shift crew composition

and commitments to Table B-1 of NUREG-0654 defines the minimum

staffing requirements for each function area.

The change to TS 6.2.2.i is administrative in nature. The

statement that reads, ``For the period ending three years after

restart from the 1993/1994 Performance Improvement Outage, the

Operations Manager will be permitted to have held a SRO license at a

Pressurized Water Reactor other than Indian Point Unit 3'', was a

relaxation of the requirements of 6.2.2.i.

These changes are administrative, and do not affect how the

plant is operated. They also follow the guidance of the Standard

Technical Specifications. Therefore, these changes will not create

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

accident previously evaluated.

3. Does the proposed amendment involve a significant reduction

in a margin of safety?

No, these TS change is administrative in nature. Removing the

statement in section 6.2.2.g that defines on shift operating

personnel and adding a new paragraph consistent with the Standard

Technical Specification is an administrative line item change that

follows NRC guidance. The current statement is not needed because TS

Table 6.2-1 defines the minimum operations shift new composition and

commitments to Table B-1 of NUREG-0654 defines the minimum staffing

requirements for each function area.

The change to TS 6.2.2.i is administrative in nature. The

statement that reads, ``For the period ending three years after

restart from the 1993/1994 Performance Improvement Outage, the

Operations Manager will be permitted to have held a SRO license at a

Pressurized Water Reactor other than Indian Point Unit 3'', was a

relaxation of the requirements of 6.2.2.i.

These changes are administrative, and do not affect how the

plant is operated. They also follow the guidance of the Standard

Technical Specifications. Therefore, these changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposed to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for Licensee; Mr. David E. Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego

County, California

Description of amendment requests: The licensee proposed changes to

Technical Specification (TS) 3.3.5 ``ESFAS Instrumentation'' to include

restrictions on operation with a channel of the refueling water storage

tank level-low input to the recirculation actuation signal (RAS) and

the steam generator pressure-low input or steam generator pressure

difference-high input to the emergency feedwater actuation signal

(EFAS) in the tripped condition. The current TS allows plant operation

in this condition indefinitely. The licensee has determined that

unacceptable consequences could result from a spurious trip of RAS or

EFAS due to operation with a channel in trip condition. The licensee

states that the proposed TS changes would improve plant operational

safety and, thereby, reduce plant risk.

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.

This change provides limits for operating with a channel of the

Refueling Water Storage Tank (RWST) Level-Low input in the

Recirculation Actuation Signal (RAS) or the Steam Generator (SG)

Pressure-Low or SG Pressure Difference (SGPD)-High input to the

Emergency Feedwater Actuation Signal (EFAS) in trip.

As a result of this change, the potential for an inadvertent

actuation of either of these two signals is reduced. The proposed

Completion Times are based on Probabilistic Risk Assessment (PRA)

considerations, and are conservative compared to the current

unlimited Completion Times.

The consequences of an inadvertent actuation of EFAS or RAS are

unaffected by this change.

Therefore, this change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

(2) Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

This proposed change provides additional time limits on

operation with a channel of the RWST Level-Low input to RAS or the

SG Pressure-Lower SGPD-High inputs to EFAS in trip. Operation in

this condition is currently allowed indefinitely. The proposed

restrictions reduce the possibility of an inadvertent actuation of

RAS or EFAS, and do not allow operation in any configuration not

currently allowed by the Technical Specifications (TSs).

Therefore, this proposed change will not create the possibility

of a new or different kind of accident from any accident that has

been 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 proposed change provides additional time limits on operation

with a channel of the RWST Level-Low input to RAS or the SG

Pressure-Low or SGPD-High inputs to RAS or EFAS in trip. The

proposed limits are conservative compared to the current

requirements, where the time limit is unrestricted. The overall

impact of the change will be [an] increase in the margin of safety.

Therefore, there will be no significant reduction in a margin of

safety as a result of this change.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

Attorney for licensee: Douglas K. Porter, Esquire, Southern

California Edison Company, 2244 Walnut Grove Avenue, Rosemead,

California 91770.

NRC Section Chief: Stephen Dembek.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,

Unit 2, Hamilton County, Tennessee

Date of application for amendments: June 7, 1999 (TS 99-09).

Brief description of amendments: The proposed amendment would

change the Sequoyah Unit 2 Technical Specification (TS) requirements by

adding a new temporary Figure 3.4-1a and temporary footnotes to TS

3.4.8, ``Specific Activity,'' Table 4.4-4, and to corresponding Bases

in order to raise the reactor coolant specific activity limit to 1.0

microcurie per milligram Dose Equivalent iodine-131 for the remainder

of Unit 2 Cycle 10 operation.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), Tennessee Valley

Authority, the licensee, has provided its analysis of the issue of no

significant hazards

[[Page 40908]]

consideration, which is presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed TS change increases the allowed reactor coolant

specific activity for iodine-131 and decreases the leakage quantity

that would be postulated to occur at the faulted steam generator

(SG) during a main steam line break (MSLB) accident. The described

changes will return these parameters to the same values under which

the plant operated prior to the implementation of TS Change 98-02

submitted on June 26, 1998. The June 26, 1998 submittal was a

voluntary change that allowed for a greater leakage quantity during

an MSLB accident as described in Generic Letter 95-05. Returning

these parameters to their previous values does not affect or

increase the probability of any accidents previously evaluated.

An increase in the consequences of an accident would not occur

because the proportional increase in reactor coolant specific

activity, while proportionally decreasing the allowable primary-to-

secondary leakage during a postulated MSLB accident to values under

which the plant was previously operated, was evaluated in [Topical

Report No.] WCAP-13990 during the establishment of the original

primary-to-secondary leak limits. No changes to the physical plant,

to the plant operation, or maintenance practices have been

implemented that would invalidate the limits defined in WCAP-13990.

The control room dose, the low population zone dose, and the

dose at the exclusion area boundary remain bounded by the acceptance

criteria of the Updated Final Safety Analysis Report. Therefore, the

proposed TS change does not result in an increase in the

consequences of an accident previously analyzed.

B. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed TS change does not alter the configuration of the

plant. The changes do not directly affect plant operation. The

change will not result in the installation of any new equipment or

systems or the modification of any existing equipment or systems. No

new operating procedures, conditions, or modes will be created by

this proposed change. SG tube structural integrity, as defined in

draft Regulatory Guide 1.121, remains unchanged. Therefore, the

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

previously evaluated is not created.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

Raising the allowed reactor coolant specific activity, while

decreasing the allowed primary-to-secondary leakage during a

postulated MSLB accident, keeps the amount of activity released to

the environment unchanged. Design basis and offsite dose calculation

assumptions remain satisfied. Therefore, the proposed change does

not result in a significant reduction in the margin of safety.

The NRC 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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Section Chief: Sheri R. Peterson.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

Date of amendment request: May 24, 1999, as supplemented by letter

dated July 9, 1999.

Brief description of amendments: The proposed license amendments

would remove several cycle-specific parameter limits from the Technical

Specifications (TSs). These parameter limits would be added to the Core

Operating Limits Report (COLR). Appropriate references to the COLR

would be inserted in the affected TSs. In addition, the core safety

limit curves would be replaced with safety limits more directly

applicable to the fuel and fuel cladding fission product barriers. The

affected Technical Specifications are: (1) TS 2.0, ``Safety Limits

(SLs),'' (2) TS 3.3.1, ``Reactor Trip System Instrumentation

Setpoints,'' (3) TS 3.4.1, ``RCS pressure temperature and flow from

Nucleate Boiling (DNB) Limits,'' and (4) TS 5.6.5, ``Core Operating

Limits Report.'' The May 24, 1999, application was previously noticed

and published in the Federal Register on June 30, 1999 (64 FR 53213).

The July 9, 1999, supplement provided proposed additional

information that would: (a) Add the Reactor Core Safety Limit figures

to the COLR, (b) clarify that the overpower N-16 setpoint remains in

the TSs, and (c) reflect NRC approval of the topical reports used to

determine the core operating limits presented in the COLR. The

supplemental information is being noticed herein to address the issue

of no significant hazards consideration.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes remove cycle-specific parameter limits from

the Technical Specifications, add them to the list of limits

contained in the Core Operating Limits Report (COLR), and revise the

Administrative Controls section of the Technical Specifications. The

proposed changes also insert the original minimum RCS flow limits

into the Technical Specifications. The changes do not, by

themselves, alter any of the parameter limits. The changes are

administrative in nature and have no adverse effect on the

probability of an accident or on the consequences of an accident

previously evaluated. The removal of parameter limits from the

Technical Specifications does not eliminate the requirement to

comply with the parameter limits.

The parameter limits in the COLR may be revised without prior

NRC approval. However, [Technical] Specification 5.6.5c continues to

ensure that the parameter limits are developed using NRC-approved

methodologies and that applicable limits of the safety analyses are

met. While future changes to the COLR parameter limits could result

in event consequences which are either slightly less or slightly

more severe than the consequences for the same event using the

present parameter limits, the differences would not be significant

and would be bounded by the requirement of specification 5.6.5c to

meet the applicable limits of the safety analysis.

Based on the above, addition of the minimum RCS flow limit into

the Technical Specifications, removal of the parameter limits the

Technical Specifications and the addition of the described limits in

the COLR, thus allowing revision of the parameter limits without

prior NRC approval, has no significant effect on the probability or

consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or

different kind of accident from any accident previously evaluated?

The proposed changes add the minimum RCS flow limit into the

Technical Specifications, remove certain parameter limits from the

Technical Specifications and add these limits to the list of limits

in the COLR, thus removing the requirements for prior NRC approval

of revisions to those parameters. The changes do not add new

hardware or change plant operations and therefore cannot initiate an

event nor cause an analyzed event to progress differently. Thus, the

possibility of a new or different kind of accident is not created.

3. Do the proposed changes involved a significant reduction in a

margin of safety?

The margin of safety is the difference between the acceptance

criteria and the associated failure values. The proposed changes do

not affect the failure values for any parameter. Though the accident

analyses, all applicable limits (i.e., relevant event acceptance

criteria as described in the NRC-approved analysis methodologies)

are shown to be satisfied; therefore, there is no impact

[[Page 40909]]

on event acceptance criteria. Because neither the failure values nor

the acceptance criteria are affected, the proposed change has no

effect on the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont.

Date of amendment request: May 26, 1999.

Description of amendment request: The licensee proposed revising

the suppression pool water temperature surveillance requirements to

specify monitoring the temperature every 5 minutes when performing

testing that adds to the suppression pool. In addition, the licensee

proposed revising the requirement to check the suppression chamber

water level and temperature from ``once per shift'' to ``daily'' and

specify that it is the average temperature that is checked.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided the NCR its analysis of the issue of no significant hazard

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's review is

presented below:

1. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment, will not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

Vermont Yankee has determined that the proposed change will not

involve a significant increase in the probability or consequences of an

accident previously evaluated. The proposed change revises the

surveillance frequency for ``once per shift'' suppression pool water

level and temperature monitoring. Additionally, the surveillance

requirement for suppression pool water temperature monitoring when

there are indications of relief valve operation that add heat to the

suppression pool is also revised. The proposed change will revise the

surveillance wording such that routine suppression pool monitoring will

be ``daily'' and an operator will verify pool temperature every 5

minutes only during testing that adds heat to the suppression pool.

Also clarified, is that the parameter being monitored is ``average''

suppression pool water temperature.

The consequence of an accident previously evaluated is not

significantly increased since the initial suppression pool water

temperature limit, which is an input valve for accident analyses, is

not changed.

The proposed change affects only surveillance requirements and does

not require any hardware or equipment modification. Equipment

operation, plant limiting conditions for operation, and accident

analyses will be unchanged. Therefore, the proposed change does not

involve a significant increase in the probability or consequences of

accidents.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment, will not create the possibility

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

evaluated.

Vermont Yankee has determined that the proposed change does not

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

accident previously evaluated. The proposed change involves revision of

Technical Specification surveillance requirements. There are no

hardware modifications or equipment changes involved and operation of

plant equipment will be unchanged. Thus, no new or different accident

precursors will be created by this change.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment, will not involve a significant

reduction in a margin or safety. VY has determined that the proposed

change does not involve a significant reduction in a margin of safety.

The proposed change involves revision of Technical Specification

surveillance requirements. There are no hardware modifications or

equipment changes involved and plant operation and accident analyses

are unchanged. The initial suppression pool water temperature limit,

which is an input value for accident analyses, is not changed.

Therefore, the proposed change will not involve a significant reduction

in the margin of safety.

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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attoney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: June 29, 1999.

Description of amendment request: The licensee proposed revising

the leak rate requirements of Technical Specifications 3.7.A.4 and

4.7.A.4 for the main steam line isolation valves. Specifically, a total

leakage rate allowable value for the sum of the four main steam lines

is proposed that is equal to four times the current individual main

steam line isolation valve leakage rate allowable value. The individual

main steam line isolation valve leakage rate allowable value is

proposed to be one half of the total leakage rate allowable value.

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 operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change does not involve a change to the plant

design or operation. As a result, the proposed change does not

affect any of the parameters or conditions that contribute to the

initiation of any accidents previously evaluated. Thus, the proposed

change cannot increase the probability of any accident previously

evaluated.

The proposed change does not affect the leak-tight integrity of

the containment structure that is designed to mitigate the

consequences of a loss-of-coolant accident (LOCA). The primary

containment must maintain functional integrity during and following

the peak transient pressures and temperatures that result from any

LOCA, thereby limiting fission product leakage following the

accident. Because the proposed change does not alter any of the

fission product lead rate assumptions used in the design basis LOCA

analysis, the analyzed consequences of the Loss of Coolant Accident

are not changed.

The control room radiological habitability analysis uses as an

input assumption main steam line leakage rate at four times the

current Technical Specifications limit. An allowable value for total

main steam line

[[Page 40910]]

leakage rate equivalent to four times the current Technical

Specifications limit for a single main steam line isolation valve is

being added by this change. Thus, there is no effect on the main

control room radiological habitability calculation.

Based on the above VY [Vermont Yankee] has concluded that the

proposed change will not result in a significant increase in the

probability or consequences of any accident previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the

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

previously evaluated.

The proposed change does not involve a change to the plant

design or operation. As a result, the proposed change does not

affect any parameters or conditions that could contribute to the

initiation of any accident. The methods of performing the tests are

not changed. No new accident modes are created. No safety-related

equipment or safety functions are altered as a result of this

change. Restating the acceptance criteria while maintaining the

assumptions of all affected calculations has no influence over nor

does it contribute to, the possibility of a new or different kind of

accident or malfunction from those previously evaluated.

Based on the above VY has concluded that the proposed change

will not create the possibility of a new or different kind of

accident from those previously evaluated.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

Restating the acceptance criteria for the main steam line

isolation valve leakage rate while maintaining the assumptions of

all affected calculations does not impact the margin of safety. The

0.6La maximum and minimum pathway leakage rate acceptance

criteria provide the previously analyzed margin of safety. The

testing method for determining the leak-tightness of the main steam

line isolation valves has not changed. The leak rate test results

are presently added to the Types B and C tests summation. The

0.6La maximum and minimum pathway leak rate acceptance

criteria and the proposed Technical Specifications requirements

provide assurance that component degradation does not impact the

assumptions used to determine, nor provide a reduction in, and the

analyzed margin of safety.

Based on the above VY has concluded that the proposed change

will not cause 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

Trobridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont.

Date of amendment request: July 12, 1999.

Description of amendment request: The amendment would revise the

value for the Safety Limit Minimum Critical Power Ratio (SLMCPR) and

delete the wording specifying these as Cycle 20 values.

Basis for proposed no significant hazards consideration

determination: As required by to CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment, will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The basis of the SLMCPR is to ensure no mechanistic fuel damages

is calculated to occur if the limit is not violated. The new SLMCPR

values preserve the existing margin to transition boiling and

probability of fuel damage is not increased. The derivation of the

revised SLMCPR for Vermont Yankee for incorporation into the

Technical Specifications, and its use to determine plant and cycle-

specific thermal limits, have been performed using NRC approved

methods. These plant-specific calculations are performing each

operating cycle and if necessary, will require future changes to

these values based upon revised core designs. The revised SLMCPR

values do not change the method of operating the plant and have no

effect on the probability of an accident initiating event or

transient.

Based on the above, Vermont Yankee has concluded that the

proposed change will not result in a significant increase in the

probability or consequences of an accident previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment, will not create the

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

previously evaluated.

The proposed changes result only from a specific analysis for

the Vermont Yankee core reload design and deletion of a cycle

specific reference for the values. These changes do not involve any

new or different method for operating the facility and do not

involve any facility modifications. No new initiating events or

transients result from these changes.

Based on the above, Vermont Yankee has concluded that the

proposed change will not create the possibility of a new or

different kind of accident from those previously evaluated.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment, will not involve a

significant reduction in a margin of safety.

The new SLMCPR is calculated using NRC approved methods with

plant and cycle specific parameters for the current core design. The

SLMCPR value remains high enough to ensure that greater than 99.9%

of all fuel rods in the core will avoid transition boiling if the

limit is not violated, thereby preserving the fuel cladding

integrity. The operating MCPR limit is set appropriately above the

safety limit value to ensure margin when the cycle specific

transients are evaluated.

As a result, Vermont Yankee has determined that the proposed

change will not result in 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: June 22, 1999 (TSCR 210).

Description of amendment request: The proposed amendments reflect

changes to the Point Beach Nuclear Plant (PBNP) Units 1 and 2 Technical

Specifications (TSs) in order to incorporate the Westinghouse 422V+

fuel assemblies into the PBNP reactor cores. 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 Point Beach Nuclear Plant in accordance with

the proposed amendments will not result in a significant increase in

the probability or consequences of an accident previously evaluated.

The accidents which are potentially affected by the parameters

and assumptions associated with this amendment have been evaluated/

analyzed and all design standards and applicable safety criteria are

met. The consideration of these changes does not result in a

situation where the design and construction standards that were

applicable prior to the change are altered. Therefore, the changed

occurring with this amendment will not result in any additional

challenges to plant equipment that could increase the probability of

any previously evaluated accident.

The proposed changes associated with this amendment do not

affect plant systems such that their function in the control of

[[Page 40911]]

radiological consequences is adversely affected. The safety

evaluation (included in Attachment 2 of this submittal) documents

that the design standards and applicable safety criteria limits

continue to be met and therefore fission barrier integrity is not

challenged. The proposed changes have been shown not to adversely

affect the response of the plant to postulated accident scenarios.

Existing system and component redundancy and operation is not being

changed by these proposed changes. These changes will therefore not

affect the mitigation of the radiological consequences of any

accident described in the FSAR [final safety analysis report].

In some cases, the results of the revised radiological analyses

are greater than those of the current FSAR analysis. In other cases,

the new and old analyses are not directly comparable because the

radiological bases for the new analyses have been upgraded to meet

more current NRC requirements. However, in all cases, the calculated

doses are well within the regulatory acceptance criteria and do not

constitute an unacceptable significant increase in consequences.

Since the actual plant configuration, performance of systems, and

initiating event mechanisms are not being changed as a result of

this evaluation, the probability or consequences of an accident

previously evaluated is not significantly increased.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The possibility for a new or different type of accident from any

accident previously evaluated is not created as a result of this

amendment. The changes described in the amendment are supported by

the analyses and evaluations described in Attachment 2 (safety

evaluation). The evaluation of the effects of the proposed changes

indicate that all design standards and applicable safety criteria

limits are met. These changes therefore do not cause the initiation

of any new or different accident nor create any new failure

mechanisms.

All equipment important to safety will continue to operate as

designed. Component integrity is not challenged. The changes do not

result in any event previously deemed incredible being made

credible. The changes do not result in more adverse conditions or

result in any increase in the challenges to safety systems.

Therefore, operation of the Point Beach Nuclear Plant in accordance

with the proposed amendments will not create the possibility of a

new or different type of accident from any accident previously

evaluated.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not involve a significant reduction in

a margin of safety.

The proposed changes do not involve a significant reduction in

the margin of safety. Existing component redundancy is not being

changed by these proposed changes. There are no new or significant

changes to the initial conditions contributing to accident severity

or consequences. The margin of safety is maintained by assuring

compliance with acceptance limits reviewed and approved by the NRC.

Since all of the appropriate acceptance criteria for the various

analyses and evaluations have been met as discussed in Attachment 2

(Safety Evaluation) of this submittal and provided for information

in Attachment 4 (PBNP FSAR Chapter 14 ``Safety Analysis'' changes

required as a result of the analyses performed for the upgraded

fuel) of this submittal, by definition there has not been a

significant reduction of any margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

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

NRC Section Chief: Claudia M. Craig.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: July 1, 1999 (TSCR 214).

Description of amendment request: The proposed amendments reflect a

change to Point Beach Nuclear Plant (PBNP) Units 1 and 2 Technical

Specification (TS) Section 15.5.4. The amendment request proposes to

remove one of the two separate methods for verifying the acceptability

of reactor fuel for placement and storage in the spent fuel pool.

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 Point Beach Nuclear Plant in accordance with

the proposed amendments will not create a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes are administrative only in that they remove

the ability to use the reference Koo method for

determining the acceptability of fuel for placement and storage in

the spent fuel pool and new fuel storage vault at the Point Beach

Nuclear Plant. Use of the remaining approved method and requirements

ensure that fuel placed or stored in the spent fuel pool and new

fuel storage vault continues to be in accordance with their

respective design and licensing basis. That is, fuel in the storage

array will continue to meet the design basis requirement that

Keff remain less than 0.95. No modifications are being

made to the spent fuel pool and its cooling system or to the new or

spent fuel storage racks. Since the design basis of the fuel and

storage racks continue to be met, operation in accordance with the

proposed amendments cannot create a significant increase in the

probability or consequences of an accident previously evaluated.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

No physical modifications are being made to the spent fuel pool

and cooling system or to the new or spent fuel storage racks. All

design basis requirements for ensuring the safe storage of fuel in

the spent fuel pool continue to be met. Therefore, operation in

accordance with the proposed amendments cannot create the

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

previously evaluated.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not create a significant reduction in a

margin of safety.

Technical Specification requirements for placing and storing

fuel in the spent fuel pool continue to ensure that the design basis

requirement, Keff for the fuel array in the spent fuel

pool and new fuel storage remains less than 0.95, is maintained. The

existing margin of safety established by this design requirement is

maintained. Therefore, operation in accordance with the proposed

amendments cannot create a 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: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for license: John H. O'Neill, Jr., Shaw, Pittman, Potts,

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

NRC Section Chief: Claudia M. Craig.

Previously Published Notice of Consideration of Issuance of

Amendments to Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

Texas Utilities Electric Company, Docket Nos. 50-445 and 50-446,

Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell

County, Texas

Date of amendment request: May 27, 1999, as supplemented by letter

dated May 28, 1999

[[Page 40912]]

Description of amendment request: The proposed amendments would add

a footnote to Technical Specification (TS) 4.8.2.1e, ``D.C. Sources-

Operating,'' which would, on a one-time basis for Unit 1 Battery

BT1ED2, allow the licensee to substitute a performance discharge test

``* * * in lieu of the battery service test required by Specification

4.8.2.1d, twice within a 60 month interval.''

Date of publication of individual notice in Federal Register: June

14, 1999. (64 FR 31881).

Expiration date of individual notice: July 14, 1999.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas.

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 1, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and 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: December 21, 1998, as

supplemented on January 28, February 18, April 2, April 15, and April

16, 1999.

Brief description of amendment: This amendment makes changes to

Facility Operating License No. DPR-35, the Technical Specifications,

and Materials License No. 20-07626-04 to reflect the transfer of the

licenses from Boston Edison Company to Entergy Nuclear Generation

Company.

Date of issuance: July 13, 1999.

Effective date; As of the date of issuance, and shall be

implemented within 30 days.

Amendment No.: 181.

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

Technical Specifications and License.

Date of initial notice in Federal Register: January 26, 1999 (64 FR

3984). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 29, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of application for amendment: March 3, 1999.

Brief description of amendment: The amendment modified Technical

Specification Table 4.6-3, ``Reactor Vessel Material Surveillance

Program Withdrawal Schedule.'' The amendment changed the withdrawal

schedule for the upcoming reactor vessel surveillance capsule pull from

approximately 15 effective full power years to approximately 18

effective full power years.

Date of issuance: July 15, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days.

Amendment No.: 182.

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

Technical Specifications.

Date of initial notice in Federal Register: May 19, 1999 (64 FR

27316). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 15, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,

St. Lucie Plant, Unit Nos. 1 and 2, St Lucie County, Florida

Date of application for amendments: December 1, 1997, supplemented

August 26, 1998.

Brief description of amendments: Revised the Technical

Specifications (TS), Appendix B, Environmental Protection Plan (Non-

Radiological), to implement the terms and conditions of the incidental

Take Statement included in the Biological Opinion issued by the

National Marine Fisheries Service, regarding endangered sea turtles.

Date of Issuance: July 2, 1999.

Effective Date: July 2, 1999.

Amendment Nos.: 162 and 103.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the TS.

Date of initial notice in Federal Register: December 31, 1997 (62

FR 68305). The supplemental letter dated August 26, 1998, provided

clarifying information that did not change the original no significant

hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated July 2, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

Nuclear Power Station, Unit No. 3, New Location County, Connecticut

Date of application for amendment: June 5, 1998, as supplemented

January 13, 1999.

Brief description of amendment: The proposed revision to the

Millstone Unit 3 licensing basis would address a recent steam generator

tube rupture (SGTR) analysis that was determined to be an unreviewed

safety question. The SGTR analyses described in the Final Safety

Analysis Report (FSAR) include an offside dose analysis and a margin to

overfill analysis. Both of the analyses have been updated. The offsite

dose analysis was updated to reflect a larger capacity for the steam

generator atmospheric dump valve (ADV) and a decrease in the operator

response time to close the ADV block valve. The

[[Page 40913]]

margin to overfill analysis was updated to reflect a new single

failure.

Date of issuance: July 2, 1999.

Effective date: As of the date of issuance and shall be implemented

within 60 days from the date of issuance.

Amendment No.: 172.

Facility Operating License Nos. DPR-49: Amendments authorizes

revisions to the FSAR,

Date of initial notice in Federal Register: July 1, 1998 (63 FR

35992).

The January 13, 1999, supplemental letter 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 July 2, 1999.

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, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Northeast Nuclear Energy Company, et al., Docket No. 50-335, Millstone

Nuclear Power Station, Unit No. 2, New Location County, Connecticut

Date of application for amendment: March 19, 1999.

Brief description of amendment: The amendment relocated Technical

Specifications Sections 3.3.3.2, ``Instrumentation, Incore Detectors,''

3.3.3.3, ``Instrumentation, Seismic Instrumentation,'' and 3.3.3.4,

``Instrumentation, Meteorological Instrumentation,'' to the Millstone,

Unit No. 2, Technical Requirements Manual. Index page V and TS Bases

have been revised to reflect the above relocations.

Dated of issuance: July 13, 1999.

Effective date: As of the date of issuance and shall be implemented

within 60 days from the date of issuance.

Amendment No.: 237.

Facility Operating License Nos. DPR-65: Amendment Revised the

Technical Specifications.

Date of initial notice in Federal Register: April 21, 1998 (64 FR

19560).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 13, 1999.

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, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of application for amendment: November 26, 1999, which was

superseded by letter dated June 1, 1998, as supplemented by letters

dated October 30, 1998, March 29, 1999, April 20, 1999, and May 28,

1999.

Brief description of amendment: These amendment would replace the

current ultimate heat sink average water temperature limit for all

combination of plant operations.

Dated of issuance: July 6, 1999.

Effective date: Both units, effective as of date of issuance and

shall be implemented within 30 days.

Amendment Nos.: 182 and 156.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 20, 1998 (63 FR

27764). The October 30, 1998, March 29, 1999, April 20, 1999, and May

28, 1999, letters 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 July 6, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilker-Barre, PA 18701.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of application for amendment: November 23, 1999.

Brief description of amendments: The amendments modified the

Susquehanna Steam Electric Station, Units 1 and 2, Technical

Specifications limiting condition for operation, 3.8.3, and

surveillance requirements, 3.8.3.1, to increase the minimum fuel oil

storage tank volume ranges.

Dated of issuance: July 7, 1999.

Effective date: Units 1 and 2, as of date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 183 and 157.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 27, 1999 (64 FR

4160).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated July 7, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

Reference, Department, 71 South Franklin Street, Wikes-Barre, PA 18701.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego

County, California

Date of application for amendments: October 17, 1997, as

supplemented March 2 and November 28, 1998.

Brief description of amendments: These amendments authorize changes

to the updated Final Safety Analysis Report (FSAR) to permit

installation of digital radiation monitors for both the containment

purge isolation and the control room isolation signals.

Date of issuance: July 12, 1999.

Effective date: July 12, 1999; implementation shall include

submission by the licensee of the revised description authorized by

these amendments with the next update of the FSAR in accordance with 10

CFR 50.71(e).

Amendment Nos.: Unit 2-154; Unit 3-145.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the FSAR.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4324).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated July 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

California, P.O. Box 19557, Irvine, California 92713.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of application for amendments: January 12,1999, as

supplemented by letters dated May 11, and June 30, 1999.

Brief description of amendments: The amendment revised Technical

Specification 3/4.7.5, Ultimate Heat Sink, by adding a new action

statement to be used in the event the plant inlet water temperature

exceeds 90 deg. F. The amendment is effective only through September

30, 1999, and is only for the current TSs. The amendment is also

limited to a maximum plant inlet water temperature of 94 deg. F. The

proposal to raise this temperature to 95 deg. F will be addressed in a

future letter.

[[Page 40914]]

Date of issuance: July 8, 1999.

Effective date: July 8, 1999, shall be implemented within 30 days

of the date of issuance.

Amendment No.: 125.

Facility Operating License No. NPF-42: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 24, 1999 ( 64

FR 9203). The May 11 and June 30, 1999, supplemental letters provided

additional clarifying information, did not expand the scope of the

application as originally noticed and did not change the staff's

original proposed no significant hazards consideration determination,

except that the licensee proposed a maximum plant inlet water

temperature of 95 deg. F. where the letters of January and May 11,

1999, proposed only 94 deg. F. The amendment is limited to a maximum

temperature of 94 deg. F.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

Dated at Rockville, Maryland, this 21st day of July 1999.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project Management Office of Nuclear

Reactor Regulation.

[FR Doc. 99-19133 Filed 7-27-99; 8:45 am]

BILLING CODE 7590-01-M