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 Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from June 9, 1997, through June 20, 1997. The

last biweekly notice was published on June 18, 1997 (62 FR 33117).

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By August 1, 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

[[Page 35847]]

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. 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.

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

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

Date of amendment request: May 6, 1997.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.7.5, ``Ultimate Heat Sink,''

and the associated bases to support steam generator replacement and to

incorporate recent Ultimate Heat Sink (UHS) design evaluations. The

replacement steam generators have a larger primary side volume which

results in a larger mass/energy release to the containment in the event

of a loss-of-coolant accident (LOCA), and a corresponding increase in

the heat load to the UHS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

TS 3/4.7.5 establishes the operating requirements for the UHS.

Operation of the UHS within its design basis ensures the following:

(1) Sufficient cooling capacity is available for continued operation

of safety related equipment during normal and accident conditions

and (2) adequate inventory is available to provide a 30-day cooling

water supply to safety related equipment. Design analyses supporting

the proposed TS changes provide full qualification of the UHS.

A loss of off site power (LOOP) coincident with a loss of

coolant accident (LOCA), designated a LOOP/LOCA, on one unit, in

conjunction with the non-accident unit proceeding to an orderly

shutdown and cooldown from maximum power using normal operating

procedures, remains the limiting design basis event for the UHS

basin temperature.

The proposed changes to the UHS Limiting Condition for Operation

for basin temperature and the number of fans running do not, in

themselves, factor into any initiating event for Updated Final

Safety Analysis Report (UFSAR) Chapter 15 accidents and,

consequently, do not increase the probability of occurrence for

these previously evaluated accidents.

The UHS plays a vital role in mitigating the consequences of any

accident or transient. The proposed changes will ensure that the

[[Page 35848]]

minimum conditions necessary for the UHS to perform its design

functions will always be met. Engineering calculations demonstrate

that the SX [essential service water] pump discharge design

temperature limit of 100 deg.F, which was assumed as an initial

input for the accident analyses, is preserved. Consequently, the

proposed changes to the number of cooling tower fans required to be

running in high speed relative to the SX pump discharge temperature

do not increase the consequences of any accident previously

evaluated.

The two unit plant trip from full power with the loss of normal

auxiliary feedwater (AF) supply source has been shown to be more

limiting than the LOOP/LOCA scenario for UHS makeup and volume

considerations.

The proposed changes to the UHS LCO for minimum basin water

level do not, in themselves, factor into any initiating event for

the UFSAR Chapter 15 accidents and, consequently, do not increase

the probability of occurrence for these previously evaluated

accidents.

The proposed changes to increase the minimum basin water levels

ensure there is a sufficient volume of water in the UHS basin at all

times. With these proposed changes, the UHS will perform its design

function for the required 30 days, and the consequences of any

accident previously evaluated are not increased.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The supporting analyses for the revised TS 3/4.7.5 do not

involve a new or different kind of accident from any accident

previously evaluated. The proposed limits on SX basin minimum water

level, maximum basin temperature, and the number of fans operating

are within the design capabilities of the UHS, and ensure that the

UHS will always be in a condition to perform its design function in

the event of an accident or transient. New and revised analyses

which support the requested TS changes ensure the full qualification

of the UHS. The UHS will not be operated in a different manner such

that the possibility of a new or different kind of accident would be

created. Consequently, these changes do not create the possibility

of a new or different kind of accident from those previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed limits on SX basin minimum water level and maximum

temperature are based on the results of new and revised design

analyses which ensure that the margin of safety is not reduced.

Required operator actions with appropriate times are incorporated

into the analyses. The new limits on temperature and volume will

ensure that, under the most limiting accident or transient scenario,

cooling water from the basin will meet the accident analyses SX

design temperature limit of 100 degrees Fahrenheit and will ensure

that adequate inventory is available to provide a 30-day cooling

water supply to safety related equipment. Therefore, the proposed

changes do not involve a significant reduction in a margin of

safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Byron Public Library District,

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

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

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: June 12, 1997.

Description of amendment request: The proposed license amendment

request would change the licensee's name from ``Duke Power Company'' to

``Duke Energy Corporation'' in the facility operating licenses for the

Catawba, McGuire, and Oconee nuclear stations as a result of a

corporate merger of Duke Power Company with PanEnergy Corporation.

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 the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. These LARs (license amendment requests) involve an

administrative change only. The Oconee, McGuire, and Catawba FOLs

(Facility Operating Licenses) are being changed to reference the new

corporate name of the licensee. No actual plant equipment or

accident analyses will be affected by the proposed changes.

Therefore, these LARs will have no impact on the possibility of any

type of accident: new, different, or previously evaluated.

(2) Will the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

No. These LARs involve an administrative change only. The

Oconee, McGuire, and Catawba FOLs are being changed to reference the

new corporate name of the licensee. No actual plant equipment or

accident analyses will be affected by the proposed changes and no

failure modes not bounded by previously evaluated accidents will be

created. Therefore, these LARs will have no impact on the

possibility of any type of accident: new, different, or previously

evaluated.

(3) Will the change involve a significant reduction in a margin

of safety?

No. Margin of safety is associated with confidence in the

ability of the fission product barriers (i.e., fuel and fuel

cladding, Reactor Coolant System pressure boundary, and containment

structure) to limit the level of radiation dose to the public. These

LARs involve an administrative change only. The Oconee, McGuire, and

Catawba FOLs are being changed to reference the new corporate name

of the licensee.

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: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730.

Attorney for licensee: Mr. Paul R. Newton, Legal Department

(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North

Carolina 28242.

NRC Project Director: Herbert N. Berkow.

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

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

Date of amendment request: June 12, 1997

Description of amendment request: The proposed license amendment

request would change the licensee's name from ``Duke Power Company'' to

``Duke Energy Corporation'' in the facility operating licenses for the

Catawba, McGuire, and Oconee nuclear stations as a result of a

corporate merger of Duke Power Company with PanEnergy Corporation.

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 the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. These LARs (license amendment requests) involve an

administrative change only. The Oconee, McGuire, and Catawba FOLs

(Facility Operating Licenses) are being changed to reference the new

corporate name of the licensee. No actual plant equipment or

accident analyses will be affected by the proposed changes.

Therefore, these LARs will have no impact on the possibility of any

type of accident: new, different, or previously evaluated.

(2) Will the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

[[Page 35849]]

No. These LARs involve an administrative change only. The

Oconee, McGuire, and Catawba FOLs are being changed to reference the

new corporate name of the licensee. No actual plant equipment or

accident analyses will be affected by the proposed changes and no

failure modes not bounded by previously evaluated accidents will be

created. Therefore, these LARs will have no impact on the

possibility of any type of accident: new, different, or previously

evaluated.

(3) Will the change involve a significant reduction in a margin

of safety?

No. Margin of safety is associated with confidence in the

ability of the fission product barriers (i.e., fuel and fuel

cladding, Reactor Coolant System pressure boundary, and containment

structure) to limit the level of radiation dose to the public. These

LARs involve an administrative change only.

The Oconee, McGuire, and Catawba FOLs are being changed to

reference the new corporate name of the licensee.

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: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina 28223-0001.

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: June 12, 1997.

Description of amendment request: The proposed license amendment

request would change the licensee's name from ``Duke Power Company'' to

``Duke Energy Corporation'' in the facility operating licenses for the

Catawba, McGuire, and Oconee nuclear stations as a result of a

corporate merger of Duke Power Company with PanEnergy Corporation.

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 the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. These LARs (license amendment requests) involve an

administrative change only. The Oconee, McGuire, and Catawba FOLs

(Facility Operating Licenses) are being changed to reference the new

corporate name of the licensee. No actual plant equipment or

accident analyses will be affected by the proposed changes.

Therefore, these LARs will have no impact on the possibility of any

type of accident: new, different, or previously evaluated.

(2) Will the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

No. These LARs involve an administrative change only. The

Oconee, McGuire, and Catawba FOLs are being changed to reference the

new corporate name of the licensee. No actual plant equipment or

accident analyses will be affected by the proposed changes and no

failure modes not bounded by previously evaluated accidents will be

created. Therefore, these LARs will have no impact on the

possibility of any type of accident: new, different, or previously

evaluated.

(3) Will the change involve a significant reduction in a margin

of safety?

No. Margin of safety is associated with confidence in the

ability of the fission product barriers (i.e., fuel and fuel

cladding, Reactor Coolant System pressure boundary, and containment

structure) to limit the level of radiation dose to the public. These

LARs involve an administrative change only. The Oconee, McGuire, and

Catawba FOLs are being changed to reference the new corporate name

of the licensee.

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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691.

Attorney for licensee: J. Michael McGarry III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036.

NRC Project Director: Herbert N. Berkow.

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

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

Connecticut

Date of amendment request: May 30, 1997.

Description of amendment request: Technical Specification (TS)

Surveillances 4.5.2.f and 4.6.2.2.b require the periodic flow testing

of the recirculation spray system pumps. The proposed amendment would

change the surveillances by replacing the pump differential acceptance

criteria with a pump acceptance curve.

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:

NNECO has reviewed the proposed revision in accordance with

10CFR50.92 and has concluded that the revision does not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

satisfied. The proposed revision does not involve [an] SHC because

the revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed changes to Technical Specification Surveillances

4.5.2.f and 4.6.2.2.b will modify the surveillance acceptance

criteria to require that each Recirculation Spray System (RSS) pump

develop a differential pressure greater than or equal to the pump

performance curve contained on Figure 3.5-1 when tested according to

the requirements of Specification 4.0.5. Because it is undesirable

to test the pumps on recirculation flow to the RWST [reactor water

storage tank], pump testing will now be performed at lower flows

than previously performed. Consistent with Specification 4.0.5, one

point on Figure 3.5-1 will be used to meet the proposed surveillance

acceptance criteria. Periodically comparing the reference

differential pressure developed at this reduced flow detects trends

that might be indicative of pump degradation. The proposed changes

are consistent with RSS pump design criteria and performing

surveillance testing does not significantly increase the probability

of an accident previously evaluated.

The proposed changes to modify the surveillance acceptance

criteria to require that each RSS pump develop a differential

pressure greater than or equal to the pump performance curve

provides the necessary assurance that the pumps will function as

required in previous evaluations and does not significantly increase

the consequence of an accident previously evaluated.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequence 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 to the surveillance acceptance criteria of

the RSS pumps does not change the operation of the Recirculation

Spray System or any of its components during normal or accident

evaluations.

Therefore, the proposed revision does 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 will change the surveillance requirements

needed to demonstrate operability for each of the RSS pumps.

Technical Specification Surveillances 4.5.2.f and 4.6.2.2.b will now

require that each pump meet its acceptance criteria in accordance

with Figure 3.5-1

[[Page 35850]]

when tested according to the requirements of Specification 4.0.5.

Figure 3.5-1 will be inserted into the Technical Specifications.

The new acceptance criteria for the RSS Technical Specification

surveillance is above the accident analysis curve and is more

restrictive than the current inservice inspection curve in the

accident analysis region. The proposed TS curve has been degraded in

accordance with the recommendations of ASME XI (American Society of

Mechanical Engineers Boiler and Pressure Vessel Code, Section XI)

for the full range of flow and will be used to meet the TS

requirements.

Therefore, the proposed revision does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve an SHC.

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, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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.

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

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

Connecticut

Date of amendment request: June 13, 1997.

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) Surveillance Requirement 4.4.1.3.3

to be consistent with the requirements of TS 3.4.1.3. Specifically, the

change would bring TS Surveillance 4.4.1.3.3 into agreement with TS

3.4.1.3 that would require at least two reactor coolant system loops to

be operable and in operation when the reactor trip system breakers are

closed during Mode 4.

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:

NNECO has reviewed the proposed revision in accordance with

10CFR50.92 and has concluded that the revision does not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

satisfied. The proposed revision does not involve (an) SHC because

the revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed change to Technical Specification Surveillance

4.4.1.3.3 is being made to bring Technical Specification

Surveillance 4.4.1.3.3 into agreement with Technical Specification

3.4.1.3 that requires at least two reactor coolant system loops to

be operable and in operation when the reactor trip system breakers

are closed during Mode 4. This requirement was incorporated into

Technical Specification 3.4.1.3 in Amendment 7. This change to the

surveillance does not alter the design, operation, maintenance or

testing of the associated systems as previously analyzed.

Therefore, the proposed revision does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

This proposed change does not introduce any new failure modes or

malfunctions, since the changes only bring Surveillance 4.4.1.3.3 in

agreement with Technical Specification 3.4.1.3. Additionally, the

proposed change does not alter the operation of the reactor coolant

system during normal or accident conditions.

Therefore, the proposed revision does 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 change to Technical Specification Surveillance

4.4.1.3.3 will reword the surveillance to ensure compliance with

Technical Specification 3.4.1.3. Technical Specification 3.4.1.3 was

changed in Amendment No. 7 to address the closure of the Reactor

Trip System breakers in Mode 4. As written, Technical Specification

Surveillance 4.4.1.3.3 does not adequately ensure compliance with

Technical Specification 3.4.1.3. This proposed change is necessary

to bring Surveillance 4.4.1.3.3 in agreement with Technical

Specification 3.4.1.3 as it was amended.

Therefore, the proposed revision does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve an SHC.

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, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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.

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

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

Minnesota

Date of amendment requests: May 7, 1997, as supplemented May 30,

1997.

Description of amendment requests: The proposed amendments would

remove from the Technical Specifications certain limitations on crane

operations in the spent fuel pool enclosure relating to spent fuel pool

special ventilation system operability.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment[s] will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Operation of the Prairie Island plant in accordance with the

proposed changes does not involve a significant increase in the

probability or consequences of an accident previously evaluated. The

proposed changes do not involve a physical modification to the

plant.

The spent fuel pool special ventilation system is provided to

mitigate the consequences of a design basis fuel handling accident

which involves dropping a spent fuel assembly directly onto a stored

spent fuel assembly. Spent fuel pool special ventilation system

performance and environmental consequences were based on the

conservative assumption that all fuel rods in one fuel assembly

fail. However, evaluation of the mechanical performance of spent

fuel stored in the spent fuel racks demonstrated that no fuel rods

fail.

The proposed changes will continue to require the spent fuel

pool special ventilation system to be operable to mitigate the

consequences of a fuel handling accident in accordance with its

original design intent. Spent fuel pool special ventilation system

operability is not required in conjunction with crane operations.

Heavy loads in the spent fuel pool enclosure are handled (1) by

single-failure-proof cranes with rigging and plant procedures which

implement Prairie Island commitments to NUREG-0612 [``Control of

Heavy Loads at Nuclear Power Plants''] or (2) over spent fuel pool

protective

[[Page 35851]]

covers as described in the Prairie Island USAR [updated safety

analysis report]. In accordance with the requirements of NUREG-0612,

use of a single-failure-proof crane with rigging and procedures

which implement the requirements of NUREG-0612 assures that the

potential for a load drop is extremely small and the effects of

heavy load drops are not considered. Spent fuel pool covers prevent

dropped loads from falling into the spent fuel pool. Thus, there are

no radiological releases resulting from handling heavy loads in the

spent fuel pool enclosure for which spent fuel pool special

ventilation system operability would be required. Therefore, these

changes do not involve a significant increase in the probability or

consequences of the fuel handling accident previously evaluated.

2. The proposed amendment(s) will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

The proposed Technical Specification changes continue to require

the spent fuel pool special ventilation system to be operable during

handling of irradiated fuel as originally designed. Heavy loads in

the spent fuel pool enclosure are handled by means which assure that

the potential for a dropped load is extremely small (through use of

single-failure-proof cranes with rigging and plant procedures which

implement Prairie Island commitments to NUREG-0612) or prevent

dropped loads from falling into the spent fuel pool (through use of

spent fuel pool protective covers as described in the USAR). Thus,

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, in themselves, do not introduce a new

mode of plant operation, surveillance requirement or involve a

physical modification to the plant.

The proposed changes do not alter the design, function, or

operation of any plant components and therefore, no new accident

scenarios are created. Therefore, the possibility of a new or

different kind of accident from any accident previously evaluated

would not be created by these amendments.

3. The proposed amendment(s) will not involve a significant

reduction in the margin of safety.

The proposed amendment(s) will continue to require the spent

fuel pool special ventilation system to operate following a fuel

handling accident as originally designed. Heavy load crane

operations in the spent fuel pool enclosure are handled (1) by

single-failure-proof cranes with rigging and plant procedures which

implement Prairie Island commitments to NUREG-0612; or (2) over

spent fuel pool protective covers as described in the Prairie Island

USAR. Provision of single-failure-proof equipment and compliance

with the other requirements of NUREG-0612 provides an equivalent

margin of safety to that which would be demonstrated by analysis of

the radiological effects of dropped loads. Use of protective covers

has been previously reviewed and approved by the NRC. Therefore,

th[ese] proposed amendment(s) (do) not involve a significant

reduction in the margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and

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

NRC Project Director: John N. Hannon.

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: May 9, 1997.

Description of amendment request: The proposed change revises the

Peach Bottom Atomic Power Station, Units 2 and 3 technical

specifications to extend the interval for replacing the primary

containment purge and exhaust valve inflatable seals.

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 TS (technical specification) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

Revising SR [surveillance requirement] 3.6.1.3.16 to replace the

inflatable seals for the Primary Containment purge and exhaust

valves from every 48 months to every 96 months will not involve a

significant increase in the probability or consequences of an

accident previously evaluated. The valves will continue to be leak

tight throughout the lifetime of the plant. This change will not

result in increased onsite or offsite radiological dose. This change

will result in reduced occupational dose exposure.

This submittal does not propose any change to the existing

requirements contained in the PBAPS [Peach Bottom Atomic Power

Station] Technical Specifications for leak testing of the Primary

Containment purge and exhaust valves per 10 CFR 50, Appendix J,

``Primary Reactor Containment Leakage Testing For Water-Cooled Power

Reactors.'' This continued testing will assure the leak tightness of

the purge and exhaust valves.

The T-ring materials (Ethylene Propylene) has been found to

withstand normal and accident thermal exposures for the design life

of the plant based on thermal aging analysis. The elastomer seat

material will provide acceptable seat tightness when exposed to a

total integrated radiation dose of 10E7 rads based on information

provided by EPRI [Electric Power Research Institute] in technical

report NP-2129, entitled ``Radiation Effects on Organic Material in

Nuclear Plants.'' The radiation dose of 10E7 rads bounds the design

basis accident dose to which these valves would be exposed. The

radiation dose these valves are exposed to during normal operation

is insignificant as compared to the accident dose. Based on this,

radiation effects from the additional exposure resulting from the

extended replacement frequency will not adversely impact the T-ring

seat material.

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

new or different kind of accident from any accident previously

evaluated.

Revising SR 3.6.1.3.16 to replace the inflatable seals for the

Primary Containment purge and exhaust valves from every 48 months to

every 96 months does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

This change does not involve any physical changes to a plant

structure, system, or component (SSC) which could act as an accident

initiator. The design, function, and reliability of the Primary

Containment purge and exhaust valves are also not impacted by this

change. This activity does not adversely influence any equipment,

which is required to be maintained operable for the prevention or

mitigation of accidents or transients. Furthermore, implementation

of the proposed changes will not adversely affect the manner in

which plant SSC are operated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

No margins of safety are reduced as a result of the proposed TS

changes. The proposed changes do not alter the intended operation of

plant structures, systems, or components utilized in the mitigation

of accidents or transients. The operating experience of these valves

and the testing performed in accordance with 10 CFR 50, Appendix J

provides a high level of confidence in the ability of these valves

to perform their intended safety function with respect to valve leak

tightness.

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

[[Page 35852]]

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

19101.

NRC Project Director: John F. Stolz.

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: May 23, 1997.

Description of amendment request: The proposed change revises the

Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Technical

Specifications (TS) to exclude the measured Main Steam Isolation Valves

(MSIVs) leakage from the total Type B and C local leak rate test (LLRT)

results.

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 TS changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Excluding the MSIV leakage from the total Type B and C LLRT

results does not involve any change in the safety function or method

of operation of any plant component, system, or structure. No new

accident initiators or failure modes are created as a result of this

change. Therefore, this change will not result in an increase in the

probability of an accident previously evaluated.

The MSIV leakage release pathway is of significance only for the

evaluation of the design basis LOCA (loss-of-coolant accident) as

described in the PBAPS, Units 2 and 3 UFSAR (updated final safety

analysis report). The doses effectively reflected in the PBAPS,

Units 2 and 3 UFSAR reflect the impact of a 0.635% Primary

Containment volume per day Primary to Secondary Containment leakage,

plus a 0.145% Secondary Containment bypass leakage to the condenser.

Since accident consequences already reflect both leakage release

pathways, the consequences of the design basis LOCA are not

increased.

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

new or different kind of accident from any accident previously

evaluated.

The MSIV's provide the means for mitigating the radiological

consequences of an accident. Revising Section 5.5.12 of the PBAPS,

Units 2 and 3 TS to exclude the measured MSIVs leakage from the

total Type B and C LLRT results has no effect on accident initiators

which lead to a new or different kind of accident. This change will

not involve any changes to plant systems, structures, or components

which could act as new accident initiators. The design, function,

and reliability of the MSIVs are also not impacted by this change.

Therefore, this change will not create the possibility of a new or

different kind of accident from any previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

No margins of safety are reduced as a result of this change to

the TS. No safety limits will be changed as a result of this TS

change. The MSIVs will continue to perform their intended safety

function. The combined dose rates from the two release paths (i.e.,

Primary to Secondary Containment leakage and Secondary Containment

bypass leakage) are unchanged as a result of this change, and are

within the limits of 10 CFR 100, and in conformance with NUREG-0737

post-accident access requirements.

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

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of amendment request: March 27, May 28, and June 4, 1997.

Description of amendment request: The proposed amendment would

revise the technical specifications (TSs) as follows:

Part 1--Boron Concentration Changes

The Cycle 2 core design for Watts Bar (WBN) will include a longer

fuel cycle and more highly enriched fuel (from 3.1 percent to 3.7

percent). To accommodate this design, the refueling water storage tank

(RWST) and accumulator boron concentrations will be increased to

provide enough boron in the sump to meet the large break loss-of-

coolant accident (LBLOCA) requirement for sump boron concentration.

This requirement is that during a LBLOCA, the core will remain

subcritical from boron provided by the emergency core cooling system

(ECCS), which takes suction from the RWST and containment sump.

The increase in RWST (TS 3.5.4) and accumulator (TS 3.5.1) boron

concentrations will be from a range of 2000-2100 ppm to 2500-2700 ppm

and from 1900-2100 to a range of 2400-2700 ppm, respectively.

Associated changes are proposed for TS Bases B 3.5.4.

Part 2--Safety Limits, Instrumentation, and Reactor Coolant System

Watts Bar has experienced hot leg temperature fluctuations,

including random spikes, which decrease the operating margin to both

the overtemperature delta temperature (OTDT) and overpower delta

temperature (OPDT) reactor trip setpoints. These fluctuations have

caused, in some cases, the plant to experience OT alarms during steady-

state operation since the temperature fluctuations reduced the

operating margin. To mitigate the temperature fluctuations and

associated alarms, the OTDT and OPDT setpoints have been enhanced to

increase the operating margin associated with these trip functions.

In addition, Watts Bar has decided to reduce the plant thermal

design flow from 97,500 gpm per loop to 93,100 gpm per loop (total of

390,000 gpm) to accommodate 10 percent steam generator tube plugging

and a 2 percent reduction in thermal design flow (RTDF).

Also, Watts Bar has decided to implement a tolerance of 0.6 deg.F

for the TS Surveillance for indicated differential temperature and 1

deg.F tolerance for the surveillance of TAVG (identified as

T prime and T double prime in the TSs). The use of this tolerance will

help to determine whether the indicated DT and TAVG should

be left as is, or rescaled during the surveillance. These tolerances

have been incorporated as biases into the uncertainty analysis for the

affected protection system functions. These functions include the OTDT,

OPDT and vessel DT equivalent to power (used in the steam generator

low-low water level trip functions). As a result of implementing these

biases into the protection system functions (and the changes to the

OTDT/OPDT setpoints and reduced TDF), the Allowable Value in the TSs

for the OTDT, OPDT and vessel DT equivalent to power functions have

been modified.

The licensee's safety evaluation has been prepared to allow for

plant operation during Cycle 2 with the revised OTDT and OPDT

setpoints, the thermal design flow of 93,100 gpm and the tolerances for

indicated differential temperature, T prime and T double prime. To

obtain sufficient departure from nucleate boiling (DNB) margin for the

OTDT/OPDT setpoint, reduced TDF and Cycle 2 design features, it was

necessary to implement the RTDP. The

[[Page 35853]]

RTDP program changes the uncertainty treatment for core power,

TAVG, pressurizer pressure, and RCS flow. These

uncertainties have been incorporated, where applicable, into the safety

analyses addressed in the Safety Evaluation.

The following TSs will be changed to incorporate the OTDT/OPDT

margin enhancement, thermal design flow of 93,100 gpm and tolerances

for indicated differential temperature, T prime and T double prime.

The Reactor Core Safety Limits (TS Figure 2.1.1-1 of the licensee's

application) have been modified to improve DNB margin. The Allowable

Values for the Vessel DT Equivalent to Power input to Steam Generator

Water Level Low-Low in the Reactor Trip System Instrumentation (Table

3.3.1-1, page 4) and Engineered Safety Feature Actuation System (ESFAS)

Instrumentation (Table 3.3.2-1, page 4), have been changed to reflect

the addition of a 0.6+F tolerance to the measurement of

indicated differential temperature.

The revised reactor core safety limits lines allow for changes in

the OTDT/OPDT reactor trip setpoints to improve operating margin. The

allowable values for these functions in the Reactor Trip System

Instrumentation (TS Table 3.3.1-1) have changed as a result of

including tolerances for indicated differential temperature, T prime

and T double prime in the uncertainty analysis. Several setpoint gains

and time constants have been modified to enhance plant operation.

Regarding the RCS Pressure, Temperature and Flow DNB Limits

(Section 3.4.1), the RCS average temperature limit has been revised to

account for the change in uncertainty from implementing RTDP. The total

RCS flow has been modified to account for the reduced thermal design

flow from 97,500 gpm to 93,100 gpm. The total flow value in the

Technical Specification includes an allowance for instrument

uncertainty.

Associated changes have been made to the following TS Bases

sections: Reactor Core Safety Limits (Section B 2.1.1); Nuclear

Enthalpy Rise Hot Channel Factor (Section B 3.2.2); Reactor Trip System

Functions OTDT, OPDT and Steam Generator Water Level Low-low (Vessel

Delta T Equivalent to Power) (Section B 3.3.1); Reactor Trip System

Functions--Reactor Coolant Flow--Low (Single Loop and Two Loops)

(Section B 3.3.1); ESFAS Instrumentation (Section B 3.3.2); RCS

Pressure, Temperature, and Flow DNB (Section B 3.4.1).

Part 3--Addition To Core Operating Limit Report Methodologies

The amendment would revise the Core Operating Limits Report (COLR)

methodologies listed in TS 5.9.5.b to add the reference to the

Westinghouse report WCAP-12610-P-A, ``Vantage + Fuel Assembly Reference

Core Report.'' The report reflects use of fuel assemblies in Cycle 2

using ZIRLO fuel rod cladding.

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:

Part 1--Boron Concentration Changes

The Nuclear Regulatory Commission has provided standards for

determining whether a significant hazards consideration exists (10 CFR

50.92 (c)). A proposed amendment to an operating license for a facility

involves no significant hazards consideration if 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;

The RWST and accumulator boron concentrations do not affect any

initiating event for accidents currently evaluated in the FSAR

[final safety analysis report]. The increased concentrations will

not adversely affect the performance of any system or component

which is placed in contact with the RWST or accumulator water. The

integrity and operability of the stainless steel surfaces in the

RWST, accumulator and affected NSSS [nuclear steam supply system]

components/systems will be maintained. The decrease in solution pH

is small and will not degrade the stainless steel. Also, the

integrity of the Class 1E instrumentation and control equipment will

be maintained since the lower sump pH, resulting from the increased

boron concentrations, is still within the applicable equipment

qualification [EQ] limits. These limits are set to preclude the

possibility of chloride induced stress corrosion cracking and assure

that there is no significant degradation of polymer materials. The

design, material and construction standards of all components which

are placed in contact with the RWST and accumulator water remain

unaffected.

For the evaluations, the consequences of an accident previously

evaluated in the FSAR will not be increased. There is no increase in

the LOCA accident consequences. The changes in the concentrations

increase the amount of boron in the sump during a LOCA. The

increased boron in the sump is sufficient to maintain the core in a

subcritical condition during a LOCA. Also, a revised hot leg

switchover time has been calculated and will be implemented in the

plant EOPs (emergency operating procedures). Thus, there will be no

boron precipitation in the core during a LOCA.

Furthermore, there is no increase in consequences of the non-

LOCA events. The concentration changes are a benefit to the SLB

(steam line break) at full power analysis due to the reduction in

power during the accident. The loss of normal feedwater event is not

sensitive to changes in the RWST and accumulator boron

concentrations. The concentration changes do not affect the

inadvertent operation of ECCS analysis since the minimum DNBR

(departure from nucleate boiling ratio) occurs at the event

initiation, and the concentration changes do not affect the analysis

trend.

Finally, the concentration changes are a benefit for the SLB M&E

(mass and energy) release and SGTR (steam generator tube rupture)

events since the increased boron increases the available shutdown

margin for these events. In addition, the increase in RWST and

accumulator boron concentrations and subsequent slight decrease in

containment sump and a spray pH does not impact the LOCA dose

evaluation since pH is not a function of radionuclide concentration.

Therefore, the present analysis remains bounding. Also, the slight

decrease in sump, core and spray fluid pH has been evaluated to not

impact the corrosion rate (and subsequent generation of Hydrogen) of

Aluminum and Zinc inside containment significantly that the present

analysis does not remain bounding. Further, the decreased sump, core

and spray fluid pH has been evaluated to not affect the amount of

hydrogen generated from the radiolytic decomposition of the sump and

core solution. In view of the preceding, it is concluded that the

proposed change will not increase the consequences of an accident

previously evaluated in the FSAR.

(2) or create the possibility of a new or different kind of

accident from any accident previously evaluated;

The changes to the RWST and accumulator concentrations do not

cause the initiation of any accident nor create any new credible

limiting single failure. The changes do not result in a condition

where the design, material, and construction standards of the RWST

and accumulators and other potentially affected NSSS components,

that were applicable prior to the changes, are altered. * * * *

The changes do not invalidate any of the accident analyses

results or conclusions. All of the safety analysis acceptance

criteria continue to be met. The changes in the concentrations

increase the amount of boron in the sump during a LOCA. The

increased boron in the sump is sufficient to maintain the core in a

subcritical condition during a LOCA. Also, a revised hot leg

switchover time has been calculated and will be implemented in the

plant EOPs. Thus, there will be no boron precipitation in the core

during a LOCA.

Furthermore, there is no possibility of a different kind of non-

LOCA event. The concentration changes are a benefit to the SLB at

full power analysis due to the reduction in power increase during

the accident. The loss of normal feedwater event is not sensitive to

changes in the RWST and

[[Page 35854]]

accumulator boron concentrations. The concentration changes do not

affect the inadvertent operation at ECCS analysis since the minimum

DNBR occurs at the event initiation, and the concentration changes

do not affect the analysis trend.

Finally, the concentration changes are a benefit for the SLB M&E

release and SGTR events since the increased boron increases the

available shutdown margin for these events.

(3) or involve a significant reduction in a margin of safety.

The changes do not invalidate any of the non-LOCA safety

analysis results or conclusions, and all of the non-LOCA safety

analysis acceptance criteria continue to be met. The margin of

safety associated with the licensing basis LBLOCA and SBLOCA (small-

break loss-of-coolant accident) analyses is not reduced as a result

of the proposed changes. Since adequate margin to the PCT (peak

cladding temperature) limit of 2200+F has been

maintained, no degradation in the margin of safety to the design

failure point (fuel melt) has been calculated. The licensing basis

containment and steam line break mass and energy releases remain

bounding, and the SGTR event acceptance criteria continue to be met.

Furthermore, the changes do not affect the safety related

performance of the RWST, accumulator or related NSSS components.

Part 2--Safety Limits, Instrumentation, and Reactor Coolant System.

The Nuclear Regulatory Commission has provided standards for

determining whether a significant hazards consideration exists (10

CFR 50.92 (c)). A proposed amendment to an operating license for a

facility involves no significant hazards consideration if 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;

The proposed changes do not result in a condition where the

design, material, and construction standards, which were applicable

prior to the changes, are altered. The revised OTDT and OPDT

setpoints do not require any hardware changes and are used for

accident mitigation. Thus, the setpoint changes do not increase the

probability of the accident.

All of the affected NSSS systems and components have been

evaluated with the TDF (thermal design flow) of 93,100 gpm. The

primary loop components (reactor vessel, reactor internals, CRDMs

(control rod drive mechanism), loop piping and supports, reactor

coolant pump, steam generator, and pressurizer) meet the applicable

structural limits with the revised TDF of 93,100 gpm and will

continue to perform their design functions. The RCCA (rod cluster

control assembly) drop time remains unaffected and the current

design core bypass flow remains valid. No additional steam generator

tubes need to be plugged to mitigate the potential for U-Bend

fatigue. Also, all of the NSSS systems will still perform their

intended design functions. The pressurizer spray flow remains above

the design value and the pressurizer relief system remains

unaffected since the TDF is lower than the current design flow and

the required pressure drop is lower. The design of the auxiliary

system components remains bounding for the revised TDF and the

corresponding changes to the NSSS thermal hydraulic parameters. In

addition, all of the NSSS/BOP (nuclear steam supply system/balance

of plant) interface systems will perform their intended design

functions. The steam generator safety valves will provide adequate

relief capacity to maintain the steam generator within applicable

design limits. The ADVs [atmospheric dump valves] will still relieve

20 percent of the maximum full load steam flow. The steam dump

system will still relieve 40 percent of the maximum full load steam

flow.

All of the applicable acceptance criteria for the accidents

described in the FSAR continue to be met. The LBLOCA analysis

currently uses a TDF of 93,100 gpm. Thus, no adjustments are

required for the LBLOCA input parameters to accommodate the TDF of

93,100 gpm. The SBLOCA has been performed with the TDF of 93,100

gpm, and the corresponding PCT is well below the 2200+F

limit. The post LOCA boron concentration and the hot leg switchover

time are unaffected. The revised thermal design procedure has been

implemented to obtain sufficient DNB margin to account for the TDF

of 93,100 gpm, the new OTDT/OPDT setpoints and the Cycle 2 design

features. All of the non-LOCA analyses have been re-analyzed or re-

evaluated and all of the applicable acceptance criteria continue to

be met.

The SLB radiological doses are unaffected and are still within

the existing licensing basis limits. The margin to overfill during

the SGTR event has been improved and the offsite doses during an

SGTR have been re-calculated and shown to be well within the

10CFR100 guidelines. The plant control systems will still provide

adequate response for the Condition 1 transients without causing a

reactor trip on OTDT and OPDT.

Finally, the changes in the tolerances for indicated

differential temperature, T prime and T double prime do not require

any hardware modifications and only require changes to the Technical

Specification Allowable Values for the OPDT and OTDT setpoints and

for the vessel DT equivalent to power functions. Thus, there is no

increase in the probability of an accident since the appropriate

Allowable Values have been modified to determine channel operability

for these functions.

(2) or create the possibility of a new or different kind of

accident from any accident previously evaluated;

The proposed changes do not cause the initiation of any accident

nor create any new limiting single failures. The OTDT and OPDT

protection functions are used for accident mitigation and do not

initiate any accidents. Also, the affected systems and components

will still perform their intended design functions.

* * *

The proposed changes do not create any new failure modes for

safety related equipment. The changes do not result in any original

design specification, such as seismic requirements, electrical

separation requirements or equipment qualification being altered.

The OTDT and OPDT setpoint changes do not require any hardware

modifications and only require adjustments to the setpoint values.

The setpoints are modeled in accident analyses which are used to

demonstrate equipment and structural qualification during a SLB.

With the setpoint changes and the TDF of 93,100 gpm, the current SLB

break M&E releases inside containment remain bounding and thus there

is no effect on the qualification of the equipment inside

containment during a SLB. The SLB M&E releases outside containment

have been re-calculated. The analysis of the impacts on equipment

qualification outside containment has been completed by generating

new temperature profiles. The application addresses and provides for

continued qualification of equipment through the normal EQ program.

Also, with the reduced TDF of 93,100 gpm, the current LOCA M&E

releases are still bounding, and thus there is no effect on the

qualification of equipment inside containment during a LOCA. The

OTDT and OPDT functions are not modeled in the LOCA analyses.

Furthermore, all of the applicable compartments and subcompartments

will maintain their integrity during the LOCA and the SLB since the

mass and energy releases for these compartments and subcompartments

remain unaffected.

In addition, the LOCA hydraulic forcing functions remain

bounding for the TDF of 93,100 gpm. Thus, the applicable NSSS

systems and components will still perform their structural functions

during a LOCA.

Finally, the changes in the tolerances for DTo, T

prime and T double prime do not require any hardware modifications

and only require changes to the Technical Specification Allowable

Values for the OPDT and OTDT setpoints and for the vessel DT

equivalent to power functions. Thus, there is no increase in the

probability of an accident different than any previously evaluated

since the appropriate Allowable Values have been modified to

determine channel operability for these functions.

(3) or involve a significant reduction in a margin of safety.

The margin of safety for the applicable safety analyses has not

been reduced. The OPDT and OTDT setpoints have been incorporated

into the affected safety analyses and all safety analysis criteria

continue to be met. All of the applicable DNB limits continue to be

met for the non-LOCA analyses. The LBLOCA input parameters do not

require adjustment for the TDF of 93,100 gpm. The SBLOCA has been

re-analyzed for the TDF of 93,100 gpm, and the SBLOCA PCT is well

below the 2200+F limit. The affected NSSS systems and

components will still meet the applicable design limits and perform

their intended safety functions with the TDF of 93,100 gpm. Also,

the SLB and LOCA M&E releases are still within the applicable

equipment qualification limits. The SGTR doses remain within the

applicable 10 CFR 100 limits, and the steam generator margin to

overfill is maintained.

Summary--Parts I and II. Based on the above, TVA has determined

that operation of

[[Page 35855]]

Watts Bar in accordance with the proposed amendment would not: (1)

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, or (3) involve a significant reduction in a margin of

safety. Therefore, operation of Watts Bar in accordance with the

proposed amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92.

Part 3--Addition to Core Operating Limit Report Methodologies

(1) involve a significant increase in the probability or

consequences of an accident previously evaluated;

The use of ZIRLOTM is already permitted by TS section

4.2.1. Accordingly, the addition of the NRC approved Westinghouse

COLR methodology reference is administrative in nature. Therefore,

there is no increase in the probability or consequences of an

accident previously evaluated.

(2) or create the possibility of a new or different kind of

accident from any accident previously evaluated;

Since the use of ZIRLOTM is already permitted by TS

section 4.2.1, the addition of the NRC approved Westinghouse COLR

methodology reference is administrative in nature. Accordingly, no

new or different kind of accident has been created from those

previously evaluated.

(3) or involve a significant reduction in a margin of safety.

The use of ZIRLOTM is already permitted by TS section

4.2.1. The addition of the NRC approved Westinghouse COLR

methodology reference is administrative in nature. Therefore, 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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

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.

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: May 2, 1997.

Brief description of amendment request: The proposed amendment

would change the main steam isolation valve (MSIV) closure time

assumption used in the main steam line break accident analysis and

referenced in the Basis for Technical Specification 4.7.

Date of individual notice in Federal Register: May 15, 1997 (62 FR

26829).

Expiration date of individual notice: June 16, 1997.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin.

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

Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: September 30, 1996, as supplemented

November 26, and December 12, 1996, February 13, March 5, April 2,

April 16, May 9, and June 3, 1997 (TSCR 192).

Description of amendment request: The proposed amendments would

change Technical Specification requirements related to the service

water system, component cooling water system, containment cooling and

iodine removal systems, auxiliary electrical systems, and the control

room emergency filtration system. The supplemental applications dated

April 2, April 16, May 9, and June 3, 1997, would eliminate separate

requirements for the component cooling water system for single-unit and

two-unit operation, revise the acceptance criteria for laboratory

testing of the control room emergency filtration system charcoal

adsorber banks from 90 percent to 99 percent, and supplement additional

information on the basis for acceptability of equipment qualification

analyses and dose assessments resulting from a loss-of-coolant

accident. The June 3, 1997, submittal requested the proposed amendments

be handled on an exigent basis based on the current schedule which

indicates that Unit 2 restart is scheduled for June 25, 1997, and Unit

1 restart is scheduled for July 1, 1997, and failure of the issuance of

the amendments by these dates would result in prevention of Point

Beach's resumption of operation.

Date of individual notice in the Federal Register: June 10, 1997

(62 FR 31636).

Expiration date of individual notice: July 10, 1997.

Local Public Document Room location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

Street, NW., Washington, DC, and at the

[[Page 35856]]

local public document rooms for the particular facilities involved.

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

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of application for amendments: June 20, 1996, as supplemented

by letters dated December 30, 1996, and March 5, 1997.

Brief description of amendments: The amendments would change the

Technical Specifications (TS) by incorporating NRC-approved thermal

limit licensing methodology in the list of approved methodologies used

in establishing the fuel cycle-specific thermal limits. In addition,

the proposed amendment will change the TS to reflect the use of Siemens

Power Corporation (SPC) ATRIUM-9B fuel for all operating Modes at

Dresden, Unit 3. The proposed amendment would also correct minor

editorial items in the TS.

Date of issuance: June 12, 1997.

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

Amendment Nos.: 160 and 155.

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

revised the licenses and the Technical Specifications.

Date of initial notice in Federal Register: April 9, 1997 (62 FR

17227). The Commission's related evaluation of the amendments is

cotained in a Safety Evaluation dated June 12, 1997.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 12, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.Consolidated

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

Generating Unit No. 2, Westchester County, New York

Date of application for amendment: March 31, 1997.Brief description

of amendment: The amendment revises Technical Specifications (TSs) to

remove the reference of Valve 863 from TS Table 3.6-1. This revision

would allow for the installation of a proposed modification for

automatic closure of Valve 863 upon receipt of a Phase A containment

Isolation signal.

Date of issuance: June 19, 1997.

Effective date: As of the date of issuance to be implemented within 30

days.

Amendment No.: 193.

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

Technical Specifications.

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

26823)The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 19, 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.

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: March 27, 1997, as supplemented by

letter dated May 6, 1997.

Brief description of amendment: The amendment changes the Technical

Specification 3/4.5.2, ``ECCS Subsystems--Modes 1, 2, and 3.'' The

proposed changes add a surveillance requirement to verify the Emergency

Core Cooling System (ECCS) piping is full of water at least once per 31

days, and clarifies wording of surveillance requirement 4.5.2.j. The

amendment also revises the TS Bases 3/4.5.2 and 3/4.5.3 to reflect

surveillance requirement.

Date of issuance: June 11, 1997.

Effective date: June 11, 1997, to be implemented within 60 days.

Amendment No.: 130.

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

Technical Specifications.

Date of initial notice in Federal Register: April 9, 1997 (62 FR

17234). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 11, 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.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,

Unit No. 1, St. Lucie County, Florida.

Date of application for amendment: December 20, 1996, and

supplemented February 13, and April 17, 1997.

Brief description of amendment: This amendment modifies the

Technical Specifications (TS) to delete a footnote associated with TS

2.1.1, ``Reactor Core Safety Limits'' which requires reactor thermal

power to be limited to 90% of 2700 Megawatts thermal for Cycle 14

operation beyond 7000 Effective Full Power Hours.

Date of Issuance: May 16, 1997.

Effective Date: May 16, 1997.

Amendment No.: 151.

Facility Operating License No. DPR-67: Amendment revised the TS.

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

2190).

The February 13, and April 17, 1997, letters provided clarifying

information that did not change the scope of the December 20, 1996,

application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 16, 1997.

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.

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

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

Minnesota

Date of application for amendments: July 28, 1995, as revised

February 21, 1997.

Brief description of amendments: The amendments revise the

Technical Specifications for the Prairie Island Nuclear Generating

Plant to allow credit for soluble boron in spent fuel criticality

analyses. The request is based on the NRC approval of the Westinghouse

Owners Group generic methodology for crediting soluble boron given in

Topical Report WCAP-14416-NP-A, ``Westinghouse Spent Fuel Rack

Criticality Analysis Methodology,'' Revision 1, November 1996.

Date of issuance: June 12, 1997.

Effective date: June 12, 1997, with full implementation within 30

days.

Amendment Nos.: 129 and 121.

Facility Operating License Nos. DPR-42 and DPR-60. Amendments

revised the Technical Specifications.

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

14464).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 12, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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: January 13, 1997, as

supplemented March 24, 1997, May 13, 1997, and May 23, 1997.

[[Page 35857]]

Brief description of amendment: The amendment revises Technical

Specifications Requirements for containment leakage testing to add

several containment isolation valves and to implement the requirements

of 10 CFR Part 50, Appendix J, Option B for performance-based primary

reactor containment leakage testing.

Date of issuance: June 17, 1997.

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

30 days.

Amendment No.: 174.

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

Technical Specifications.

Date of initial notice in Federal Register: March 19, 1997 (62 FR

13173).

The March 24, May 13, and May 23, 1997, supplemental letters

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 June 17, 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.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of application for amendments: January 31, 1997.

Brief description of amendments: The amendments revise Technical

Specification 3/4.6.1.5, and its associated Bases section, to ensure

that a representative average containment air temperature is measured.

Date of issuance: June 13, 1996.

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

implemented within 60 days.

Amendment Nos. 195 and 178.

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

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

11497).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 13, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

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: August 22, 1996, as

supplemented March 28, 1997.

Brief description of amendments: Revise Technical Specifications

(TS) 3.6.5 and associated Bases to lower the minimum TS ice basket

weight. Also extend the chemical analysis surveillance interval for the

ice condenser ice bed from 12 months to 18 months.

Date of issuance: June 10, 1997.

Effective date: June 10, 1997.

Amendment Nos.: 224, 215.

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

revise the TS.

Date of initial notice in Federal Register: April 23, 1997 (62 FR

19835).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 10, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: April 22, 1997, as supplemented

on May 15, and June 2, 1997. The April 22, 1997, submittal superseded a

previous submittal on this subject dated September 6, 1996 (61 FR

53769), as supplemented on October 30, October 31, November 7, November

15, and November 27, 1996, and January 23 and January 29, 1997.

Brief description of amendment: The amendment revises TS Section

4.2.b, ``Steam Generator Tubes,'' and its associated Basis, by allowing

a laser-welded repair of Westinghouse hybrid expansion joint (HEJ)

sleeved steam generator tubes.

Date of issuance: June 7, 1997.

Effective date: June 7, 1997.

Amendment No.: 135.

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 7, 1997 (62 FR

24988).

The May 15, and June 2, 1997, submittals provided supplemental

information that did not change the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 7, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

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

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: April 24, 1997, as supplemented

on May 15 and 28, and June 5, 1997.

Brief description of amendment: The amendment revises 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 are also 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.''

Date of issuance: June 7, 1997.

Effective date: June 7, 1997.

Amendment No.: 134.

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 7, 1997 (62 FR

24989).

The May 15 and 28, and June 5, 1997, submittals provided

supplemental information that did not change the initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 7, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

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

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: April 28, 1997, as supplemented

on May 19, 1997.

Brief description of amendment: The amendment establishes a new

design basis flow rate for the auxiliary feedwater (AFW) pumps

consistent with the assumptions used in the reanalysis of the limiting

design basis event for the

[[Page 35858]]

AFW system. The Basis for TS 3.4.b, ``Auxiliary Feedwater System,'' has

been revised to reflect the change in AFW flow and to clarify the

requirements for the AFW cross-connect valves.

Date of issuance: June 7, 1997.

Effective date: June 7, 1997.

Amendment No.: 133.

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 7, 1997 (62 FR

24977).

The May 19, 1997, submittal provided clarifying information that

did not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated June 7, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

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

Dated at Rockville, Maryland, this 25th day of June, 1997.

For the Nuclear Regulatory Commission.

Jack W. Roe,

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

Reactor Regulation.

[FR Doc. 97-17140 Filed 7-1-97; 8:45 am]

BILLING CODE 7590-01-P