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

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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 December 21, 1995, through January 4, 1996.

The last biweekly notice was published on January 3, 1996 (61 FR 174).

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 Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, 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 February 21, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC and at the local public

document room for the particular facility involved. If a request for a

hearing or petition for leave to intervene is filed by the above date,

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

Board Panel, will rule on the request and/or petition; and the

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) The nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any

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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, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington DC, by the above

date. Where petitions are filed during the last 10 days of the notice

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

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

Nuclear Regulatory Commission, Washington, DC 20555, 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units

Nos. 1, 2, and 3, Maricopa County, Arizona.

Date of amendments request: December 19, 1995

Description of amendments request: The proposed amendments would

allow the implementation of the recently approved Option B to 10 CFR

Part 50, Appendix J. This new rule allows for a performance-based

option for determining the test frequency for containment leakage rate

testing. The proposed amendment would modify Technical Specifications

(TS) 1.7, 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, and 3/4.6.3 and the Bases of

TS 3/.6.1.2. It would also create a new TS 6.16.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed Technical Specification (TS) changes will result in

generally increased intervals between containment leakage rate tests

determined through a performance based approach. The interval

between such tests are not related in any way to conditions which

cause accidents. Plant structures, systems, and components will not

be operated in a different manner as a result of the proposed TS

change, therefore, the proposed changes will not increase the

probability of an accident previously evaluated.

Containment leakage may result from accidents which are

evaluated in the Updated Final Safety Analysis Report. The proposed

TS changes may result in a small, but acceptable, increase in post-

accident containment leakage. This increase is calculated as a

statistical expectation using the probability that leakage through a

penetration will exceed the administrative limit and through the

increased time needed to detect such excess leakage. NUREG-1493,

which is the technical basis for 10 CFR Part 50, Appendix J, Option

B, contains a detailed evaluation of the expected leakage and its

consequences.

The increased risk due to the lengthening of the intervals

between Type A, B, and C leakage rate tests is also evaluated in

NUREG-1493. Using a statistical approach, NUREG-1493 determined that

the increase in expected dose to the public, resulting from

extending the testing interval, is extremely small. NUREG-1493

concluded that the small increase is justifiable due to the benefits

which accrue from interval extension. The primary benefit is the

reduction in occupational exposure. The reduction, on a per person

basis, is orders of magnitude greater than the marginal, potential

increase in dose to the public. The reduction in occupational

exposure is a real reduction, while the small increase in dose to

the public is statistically derived using conservative assumptions.

Therefore, the proposed change does not significantly increase the

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change does not create the possibility of a new or

different kind of accident from any previously evaluated. The

proposed change only incorporates the performance based approach

authorized in the new Option B to Appendix J of 10 CFR Part 50. The

interval extensions allowed, through this approach, do not have the

potential for creating the possibility of new or different kinds of

accidents from those previously evaluated. Plant structures,

systems, and components will not be operated in a different manner

as a result of the TS change and, therefore, will not introduce any

new or different failure modes or initiators.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed Technical Specification does not alter the

allowable containment leakage rate. The proposed change replaces the

current, prescriptive testing requirements with a new performance

based approach for establishing the testing intervals therefore, the

proposed change does not involve a significant reduction in the

margin of safety.

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

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

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

amendments request involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: William H. Bateman.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs

Nuclear Power Plant, Unit No. 1, Calvert County, Maryland.

Date of amendment request: December 21, 1995.

Description of amendment request: The proposed amendment would

revise the Calvert Cliffs Nuclear Power Plant,

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Unit No. 1, Technical Specifications (TSs). The requested change would

allow the use of cladding materials other than Zircaloy or ZIRLO. A

Temporary Exemption was issued on November 28, 1995 (60 FR 62483)

approving the loading of four (4) lead fuel assemblies (LFAs) into the

Unit No. 1 reactor vessel during cycles 13, 14, and 15. The technical

basis for the Exemption, which is the same basis for the requested TS

amendment, was provided in the Baltimore Gas and Electric Company (BGE)

submittal dated July 13, 1995. The submittal addressed the safety

significance of operating with 4 LFAs in Calvert Cliffs Nuclear Power

Plant, Unit No. 1, reactor vessel during cycles 13, 14, and 15.

Specifically, BGE proposes to add a statement to TS 5.2.1, ``Fuel

Assemblies,'' indicating, for Cycles 13, 14, and 15 only, advanced

cladding material may be used in 4 lead test assemblies as described in

a approved Temporary Exemption dated November 28, 1995.

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. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed change is to add an approved temporary exemption to

the Unit 1 Technical Specifications allowing the installation of

four lead fuel assemblies. These four assemblies use an advanced

cladding material which is not specifically permitted by existing

regulations or Calvert Cliffs' Technical Specifications. A temporary

exemption to allow the installation of these assemblies was approved

on November 28, 1995. The addition of this approved temporary

exemption to Technical Specification 5.2.1 is simply intended to

allow their installation under the provisions of the temporary

exemption. The license amendment is effective only as long as the

exemption is effective. The addition of the approved temporary

exemption to Unit 1 Technical Specification 5.2.1 does not change

the probability or consequences of an accident previously evaluated.

Therefore, this change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

The proposed Technical Specification change adds an approved

temporary exemption to Technical Specification 5.2.1 for Unit 1.

This change does not add any new equipment, modify any interfaces

with existing equipment, change the equipment's function, or change

the method of operating the equipment. The proposed change does not

affect normal plant operations or configuration. Since the proposed

change does not change the design, configuration, or operation, it

could not become an accident initiator.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. Would not involve a significant reduction in a margin of

safety.

The proposed change is to add an approved temporary exemption to

the Unit 1 Technical Specifications allowing the installation of

four lead fuel assemblies. These four assemblies use an advanced

cladding material which is not specifically permitted by existing

regulations or Calvert Cliffs' Technical Specifications. A temporary

exemption to allow the installation of these assemblies was approved

on November 28, 1995. The addition of this approved temporary

exemption to Technical Specification 5.2.1 is simply intended to

allow their installation under the provisions of the temporary

exemption. The license amendment is effective only as long as the

exemption is effective. This amendment does not change the margin of

safety by adding a reference to an approved, temporary exemption to

the Technical Specifications.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

this review, it appears that the three standards of 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: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: Ledyard B. Marsh.

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: December 7, 1995.

Description of amendment request: The proposed amendments will

remove the Technical Specification (TS) requirements for the main

feedwater pump discharge pressure switch input to the Anticipatory

Reactor Trip System (ARTS) and the Emergency Feedwater System (EFDW).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated:

No. The accidents addressed within the Oconee Final Safety

Analysis Report (FSAR) have been reviewed with respect to this

proposed Technical Specification amendment request. The probability

or consequences of any accident previously evaluated is not

significantly increased by the proposed amendment. Emergency

Feedwater is required for the mitigation of some accidents and the

availability of this system will be unaffected by this proposed

revision. Both manual and automatic actuation of the EFDW system on

a loss of main feedwater will remain.

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

accident from any kind of accident previously evaluated:

No. This amendment eliminates a portion of the automatic

actuation circuitry for EFDW and ARTS. This circuitry removal does

not create the possibility of a new or different kind of accident as

the design of the circuitry is to sense a loss of main feedwater and

supply a signal for the initiation of ARTS and EFDW. A loss of main

feedwater signal will continue to be supplied to ARTS and EFDW;

however, this loss will be sensed by low hydraulic oil pressure on

the Main Feedwater Pumps (ARTS and EFDW) and low steam generator

level (EFDW only) rather than by a low Main Feedwater Pump discharge

pressure. Since a loss of Main Feedwater will continue to be

recognized, the system will continue to function as before. Hence,

no new or different accidents will be created.

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

No. The margin of safety will not be significantly reduced as an

actuation signal to ARTS and EFDW will continue to be generated by a

loss of Main Feedwater. Consequently, ARTS and EFDW will continue to

perform the safety function required for accident mitigation.

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.

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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 amendment request: November 22, 1995.

Description of amendment request: The proposed amendments will

upgrade existing TS [Technical Specification] 3/4.4.6.1 for the Reactor

Coolant System Leakage Detection Instrumentation by adapting the

Standard Technical Specifications for Combustion Engineering Plants

(NUREG-1432), Specification 3.4.15, to both St. Lucie units. The

proposal is consistent with the NRC Final Policy Statement on Technical

Specifications Improvements (58 FR 39132).

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 or consequences of an accident previously evaluated.

The Reactor Coolant System (RCS) Leakage Detection

Instrumentation Systems are not accident initiators, and their

operational status is not a consideration in determining the

probability of occurrence of accidents previously evaluated. The

proposed revision to the related Limiting Condition for Operation

(LCO) 3/4.4.6.1 does not involve a change to the configuration or

method of operation of any equipment that is used to mitigate the

consequences of an accident, nor do the changes alter any

assumptions made involving initial plant conditions in the safety

analyses. Therefore, operation of the facility in accordance with

the proposed amendment would not involve a significant increase in

the probability 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 accident previously evaluated.

The proposed revision to LCO 3/4.4.6.1 is administrative in

nature and will not result in a change to the physical plant or the

modes of plant operation defined in the Facility License. The

revision does not involve the addition or modification of equipment

nor does it alter the design of plant systems. Therefore, 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 accident previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The RCS Leakage Detection Systems are designed to provide

diverse methods to assist in the detection and location of

unidentified leakage that may be associated with potential pressure

boundary degradation. These systems provide no equipment control or

accident mitigation functions, and are not associated with the

safety margin established for protection from analyzed Loss of

Coolant Accidents. The proposed revision to LCO 3/4.4.6.1 does not

alter the basis for any technical specification that is related to

the establishment of, or the maintenance of, a nuclear safety

margin; and simply adapts the corresponding and previously reviewed

specification from the Standard Technical Specifications for

Combustion Engineering Plants, NUREG-1432, to the St. Lucie units.

Therefore, operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

Based on the above discussions and the supporting Evaluation of

Technical Specification changes, FPL has determined that the

proposed license amendment involves no significant hazards

consideration.

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 request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

Attorney for licensee: Harold F. Reis, Esquire, Newman and

Holtzinger, 1615 L Street, NW., Washington, DC 20036.

NRC Project Director: David B. Matthews, Director.

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

Nuclear Generating Station, Ocean County, New Jersey.

Date of amendment request: December 5, 1995.

Description of amendment request: The proposed amendment revises

the submittal date in the Annual Exposure Data Report which brings

Oyster Creek into conformance with 10 CFR 20.2206 and relaxes an overly

restrictive administrative requirement.

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:

. . . The changes do not:

1. Involve a significant increase in the probability or the

consequence of an accident previously evaluated.

This change is administrative in nature and has no effect on the

operation of the plant. This change will not increase the

probability or consequence of an accident previously evaluated.

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

from any previously evaluated.

Operation of the facility in accordance with this proposed

change will not create the possibility for an accident or

malfunction of a different type from any accident previously

evaluated. The proposed amendment does not modify any system

(component) operation or maintenance activity. The facility will

continue to be operated within the limits of existing accident

analysis and margins of safety.

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

This change brings the submittal date for the Annual Exposure

Data Report into conformance with 10 CFR 20.2206 and relaxes an

overly restrictive administrative requirement. Since the proposed

change does not alter any system hardware or design basis, the

margin of safety is not reduced.

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: Ocean County Library,

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

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

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

NRC Project Director: Phillip F. McKee.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa.

Date of amendment request: November 15, 1995.

Description of amendment request: The proposed amendment would

revise the requirements for the End of Cycle Recirculation Pump Trip

logic to match more closely the assumptions applicable to the turbine

trip events for which it was installed. The surveillance requirements

are also proposed to be revised, based on those same assumptions.

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 Technical Specification (TS) amendment will not

significantly increase the probability or consequences of any

previously evaluated accidents. The [End of Cycle] (EOC)

[recirculation pump trip] RPT system was installed to preclude

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violation of reactor fuel limits, and the system will be preserved for

that purpose. In the event that system is not available, an

operating penalty will be imposed on the [Minimum Critical Power

Ratio] MCPR limit to assure sufficient margin to the limit to

preclude fuel damage during the postulated turbine trip events.

The change to the ``Minimum Operable Channels per Trip System''

will assure that inputs monitoring both the turbine control valve

fast closure and the turbine stop valve closure will be available to

initiate (EOC)RPT.

The change to the ``Applicable Operating Mode'' is an editorial

change which reflects the existing hardware bypass.

The change to Action 81 in TS Table 3.2-G will assure that when

the (EOC)RPT system does not meet the minimum TS availability

requirements, the [safety limit minimum critical power ratio] SLMCPR

will not be challenged. By imposing an [operating limit minimum core

power ratio] OLMCPR penalty for continued operation, the fuel

thermal limits will not be challenged, since the (EOC)RPT system was

installed to accomplish the same goal. No increase in the

consequences of the turbine trip events will result from this

change. The OLMCPR penalty is dependent on cycle-specific parameters

and will therefore be included in the cycle-specific [Core Operating

Limits Report] COLR.

The change to the surveillance interval results in (EOC)RPT

logic channel functional tests being performed once per quarter

instead of once per month. The change also revises the allowed out-

of-service time (AOT) for testing from two hours to six hours. These

changes are consistent with the Improved Standard Technical

Specifications, NUREG-1433, Revision 1. The (EOC)RPT is initiated by

instruments common to the Reactor Protection System (RPS) (i.e.,

turbine stop valve closure and turbine control valve fast closure).

The surveillance interval and AOT changes for these instruments were

evaluated in ``Technical Specification Improvement Analysis for BWR

Reactor Protection System,'' NEDC-30851P-A, March 1988, for the RPS

function. Although the (EOC)RPT functions were not explicitly

identified in that document, these changes can be considered bounded

by that analysis. The basis for this conclusion is similar to the

basis established for the control rod block instrumentation common

to the RPS, as documented in ``Technical Specification Improvement

analysis for BWR Control Rod Block Instrumentation,'' NEDC-30851P-A,

Supplement 1, October 1988. Failure of the (EOC)RPT function could

potentially lead to exceeding the SLMCPR, similar to the

consequences of an unmitigated rod withdrawal error. The slight

increase in risk of a SLMCPR violation due to extending (EOC)RPT

surveillance interval and AOT is offset by the same benefits

associated with the similar approved surveillance interval and AOT

for the RPS. Both the above referenced reports have been approved

for application at the DAEC via TS Amendment 193, dated April 14,

1993.

The changes to the ``Operating Modes for which Surveillance

Required'' are clarifications and will result in a more efficient

utilization of resources. By stating that the surveillance applies

only when the (EOC)RPT system is OPERABLE, the surveillances will

not be performed needlessly. During the early part of an OPERATING

cycle, the (EOC)RPT is not required to mitigate a turbine trip, and

therefore, may be bypassed. At the time when the (EOC)RPT is assumed

to be OPERABLE pursuant to the analysis, it will be made OPERABLE

unless accepting the penalty on the OLMCPR is preferable. The result

of the proposed change will still be that the (EOC)RPT is

demonstrated OPERABLE at any time when it is required.

The change to the acceptance criteria for response time testing

reflects a recent review of the analytical assumptions and the

testing methodology. The (EOC)RPT is assumed to interrupt power to

the recirculation pump motor within 175 milliseconds after

initiation of either turbine stop valve closure or turbine control

valve fast closure. The response time test only measures a portion

of the complete trip (the rest was measured as part of start-up

testing). The portion measured is dependent on which trip input is

being tested. The turbine control valve closure is sensed by a

pressure switch monitoring the hydraulic fluid controlling the valve

and therefore has no delay between valve motion and initiation of

the (EOC)RPT logic. The turbine stop valve closure is sensed by

position switch. Since this switch is set to initiate (EOC)RPT at

10% valve closed, there is a brief delay between the beginning of

valve motion and initiation of the (EOC)RPT logic. The respective

proposed response time tests account for these differences, as

described in the footnotes on TS page 3.2-36, and demonstrate that

the measured portions of the action are within allowed time periods.

None of the proposed changes will significantly increase the

probability of any accident previously evaluated because the

(EOC)RPT is not an initiator of any of those events. None of the

proposed changes will significantly increase the consequences of an

accident because the (EOC)RPT system serves to prevent a turbine

trip event from exceeding the fuel SLMCPR, and it will continue to

perform in that capacity at any time when it is required to assure

margin to the SLMCPR.

2. The proposed changes will not add a new or different kind of

accident because the plant will not be operated in a different way.

By allowing the implementation of a penalty on OLMCPR in lieu of

reducing reactor power, the risk of a plant transient is reduced.

Similarly, the surveillance interval and AOT extensions will also

result in fewer plant power reductions for testing.

The (EOC)RPT initiates a trip of the recirculation pumps and any

TS change affecting that system cannot result in an effect on any

system other than those pumps. Consequently, no new accidents are

postulated as a result of this proposed change.

3. The proposed change will not result in a significant

reduction in any margin of safety. The (EOC)RPT performs to assure

adequate margin to the SLMCPR. The proposed change will preserve

that function and require that additional margin to the SLMCPR be

imposed for those times when the (EOC)RPT is not OPERABLE. The other

changes are proposed because they assure correct (EOC)RPT function

(inputs and response times).

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: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401.

Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan,

Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: Gail H. Marcus.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.

50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois.

Date of amendment request: December 14, 1995.

Description of amendment request: The proposed amendment would

modify Technical Specification 3.4.2, ``Flow Control Valves (FCVs),''

by deleting the requirement to verify that the average rate of movement

of each reactor recirculation system FCV is limited to less than or

equal to 11% per second in the opening and closing directions

(Surveillance Requirement 3.4.2.2).

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 Clinton Power Station (CPS) Updated Safety Analysis

Report (USAR) evaluates three specific events related to operation

of the reactor recirculation flow control valves (FCVs). The impact

of the proposed change on each of these events is discussed below.

The loss of coolant accident (LOCA) analysis described in USAR

Section 6.3.3.7.2 assumes that the FCVs fail ``as is'' in the event

of a LOCA. This feature is assured by electronic interlocks in the

FCV control circuitry and periodically verified as required by

Technical Specification (TS) Surveillance Requirement (SR) 3.4.2.1.

The design of these interlocks and the testing requirements are not

affected by this proposed change.

The Recirculation Flow Controller Failure--Decreasing Flow

transient analyses are described in USAR Section 15.3.2, and the

Recirculation Flow Controller Failure--Increasing Flow transient

analyses are described in USAR Section 15.4.5. Since the

[[Page 1631]]

control circuitry for the FCVs has been modified such that the

capability to operate in a master controller mode has been

eliminated, each FCV is now individually controlled, and the

possibility that a single failure could affect operation of more

than one FCV has also been eliminated. As a result, fact closure and

fast opening of both FCVs are no longer postulated for CPS. Thus,

the surveillance (SR 3.4.2.2) associated with verifying that FCV

movement is within the assumptions of the analyses for fast closure

and fast opening of both FCVs can be deleted.

With respect to fast closure and fast opening of individual

FCVs, the modification performed during the fifth refueling outage

only affected the electronic master control of the FCVs and did not

affect the hydraulic limitations of the FCVs. Conservative analyses,

component testing, and the Initial Startup Test program provide

confidence that individual FCV stroke rates assumed in the transient

analyses will not be exceeded over the life of the plant. These

analyses and conditions are sufficient to assure individual FCV

stroke rates are adequately limited without the periodic performance

of a specific test.

In addition to the above, the modification did not add any new

failure modes to the design of the individual FCV controllers. In

fact, failure modes associated with misoperation of the common

master controller have been eliminated from the control circuit

design. The modification did not alter any of the features

associated with initiators of any LOCA or features which assure that

the FCVs fail ``as is'' in the event of a LOCA.

Based on the above, Illinois Power (IP) has concluded that this

request does not increase the probability or the consequences of any

accident (or transient) previously evaluated.

(2) USAR Sections 15.3.2 and 15.4.5 describe the plant response

to malfunctions of FCV control failures, and USAR Section 6.3.3.7.2

describes the assumptions made with respect to FCV failures and

their impact on the LOCA analysis. The proposed change (and the

associated modification prompting the proposed change) does not

affect any other structures, systems, or components beyond the FCVs.

All associated failure modes thus remain within the scope of the

failure modes previously considered. As a result, IP has concluded

that the proposed change cannot create the possibility of an

accident not previously evaluated.

(3) This request does not involve any change to the requirements

or design associated with initiation or mitigation of a LOCA. The

consequences of transients associated with fast closure and fast

opening of reactor recirculation system FCVs are bounded by the

consequences of other transient events and thus are not utilized in

establishing plant operating limits. Although the control circuitry

for the FCVs was modified during the fifth refueling outage, that

modification did not affect the hydraulic failure modes of the FCVs.

Further, the modification did not add any new failure modes to the

design of the individual FCV controllers. In fact, failure modes

associated with misoperation of the common master controller have

been eliminated from the control circuit design. As a result,

assumed FCV operation during analyzed accidents and transients has

not been altered. Conservative analysis, component testing, and the

Initial Startup Testing program have confirmed that the FCV velocity

assumed in the transient analyses will not be exceeded over the life

of the plant. Thus, verification of rate of FCV movement in the

opening and closing directions need not be performed by periodic

testing and SR 3.4.2.2 can be deleted without resulting in 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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727.

Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and

Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606.

NRC Project Director: Gail H. Marcus.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.

50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois.

Date of amendment request: December 14, 1995.

Description of amendment request: The proposed amendment would

consist of several changes to the instrumentation sections of the

Clinton Power Station Technical Specifications. The proposed changes

are required due to engineering reanalyses or plant modifications. The

affected instrumentation includes: (1) steam line flow high channels

for the Reactor Core Isolation Cooling (RCIC) System, (2) ambient

temperature channels in the Residual Heat Removal (RHR) System heat

exchanger rooms, (3) reactor vessel pressure channels that provide a

permissive for operation of the shutdown cooling mode of the RHR

system, and (4) RCIC storage tank water level instrument channels.

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) None of the proposed changes involve a significant increase

in the probability or consequences of any accident previously

evaluated.

The changes to Table 3.3.6.1-1 Functions 3.a and 3.i are

administrative in nature and bring the technical specifications (TS)

into conformance with the Clinton Power Station (CPS) as-built

design. The reactor core isolation cooling (RCIC) system steam line

flow trip Function names have been changed to reflect the

elimination of the residual heat removal (RHR) steam condensing

mode. However, these trips have not been physically altered and thus

will continue to operate as before. As a result of the elimination

of the RHR steam condensing mode, the possibility of a leak in the

RCIC steam supply resulting in an increase in the RHR heat exchanger

room ambient temperature has also been eliminated. Accordingly, the

RHR ambient temperature isolation trip is changed to only isolate

the RHR system when the RHR heat exchanger room ambient temperature

setpoint is exceeded. The Shutdown Cooling System Reactor Vessel

Pressure--High function is provided to isolate the shutdown cooling

portion of the RHR system since this piping is designed for

pressures lower than rated reactor vessel pressure. This interlock

(RHR cut in permissive) is provided only for equipment protection to

prevent an intersystem LOCA scenario and credit for the interlock is

not assumed in the accident or transient analysis in the Updated

Safety Analysis Report (USAR).

The proposed change to the setpoint (Allowable Value) is

conservative with respect to considerations for shutting the RHR

shutdown cooling motor-operated valves and providing

overpressurization protection for the low pressure RHR shutdown

cooling system piping. With respect to the RCIC storage tank water

level setpoints, no accident or transient analysis takes credit for

the volume of water in the RCIC storage tank. In addition, the

setpoint (Allowable Value) has been changed to ensure RCIC system

operation is not adversely affected by a low level in the storage

tank.

The proposed changes do not affect any of the parameters or

conditions that contribute to initiation of any accidents previously

evaluated. In addition, the proposed changes do not affect the

ability of the associated instrumentation to operate as assumed in

the safety analyses. As a result, the proposed changes will not

result in a significant increase in the consequences of any accident

previously evaluated.

(2) None of the proposed changes create the possibility of a new

or different kind of accident from any accident previously

evaluated. The proposed changes for RHR/RCIC Steam Line Flow--High

[are] administrative in nature and will simply make this item

description accurate. The RCIC steam supply line no longer supplies

any steam to the RHR heat exchanger room. As a result, the

associated isolation of the RCIC system is no longer required. The

Shutdown Cooling System Reactor Vessel Pressure - High function will

still perform as designed. The RCIC Storage Tank Level - Low trip

will continue to perform in accordance with design. None of the

above listed changes will introduce any new failure modes or changes

in plant operation.

As a result, the proposed changes cannot create the possibility

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

evaluated.

(3) None of the proposed changes involve a significant reduction

in a margin to safety.

[[Page 1632]]

The proposed changes for RHR/RCIC Steam Line Flow--High do not involve

a significant reduction in a margin of safety because the change is

administrative in nature and will simply make the descriptions

accurate and consistent with completed modifications. The

elimination of RCIC system isolation in response to a high RHR room

ambient temperature is no longer required due to the elimination of

the RHR steam condensing mode. Removing the RHR room ambient

temperature isolation of the RCIC will reduce the number of

unnecessary isolations of RCIC. The Shutdown Cooling System Reactor

Vessel Pressure - High function will still perform as designed. The

proposed change to the setpoint (Allowable Value) is conservative

with respect to considerations for shutting the RHR shutdown cooling

motor-operated valves and providing overpressurization protection

for the low pressure RHR shutdown cooling system piping. The

Allowable Value for the RCIC Storage Tank Level - Low Function has

been changed to be more conservative to ensure the RCIC and HPCS

systems will perform their system safety function. No credit is

taken for the volume in the RCIC storage tank for the HPCS or RCIC

systems in performing their safety-related functions.

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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and

Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606

NRC Project Director: Gail H. Marcus.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan.

Date of amendment requests: December 19, 1995 [AEP:NRC:1215B]

Description of amendment requests: The proposed amendments would

modify the technical specifications to replace the existing scheduling

requirements for overall integrated and local containment leakage rate

testing with a requirement to perform the testing in accordance with 10

CFR Part 50, Appendix J, Option B. Option B allows test scheduling to

be adjusted based on past performance.

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:

Criterion 1

This amendment request does not involve a significant increase

in the probability or consequences of an accident previously

evaluated because the proposed changes to the T/Ss do not affect the

assumptions, parameters, or results of any UFSAR [updated final

safety analysis report] accident analysis. The proposed changes do

not change the acceptance criteria for containment leakage limits

and do not modify the response of the containment during a design

basis accident. The proposed amendment does not add or modify any

existing equipment. The proposed Types A, B, and C testing schedules

will be consistent with Appendix J Option B to 10 CFR 50 which was

developed based on analytical efforts documented in NUREG-1493

[Performance-Based Containment Leak-Test Program]. The analysis

confirms previous observations of insensitivity of population risks

from severe reactor accidents to containment leakage rates. Based on

these considerations, it is concluded that the changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

Criterion 2

The proposed changes do not involve physical changes to the

plant or changes in plant operating configuration. The proposed

changes only remove the restrictive schedular requirements for

conducting Types A, B, and C testing from the T/Ss and substitute

the schedule specified in Appendix J Option B to 10 CFR 50 and

Regulatory Guide 1.163 [Performance-Based Containment Leak-Test

Program]. Thus, it is concluded that the proposed changes do not

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

any accident previously evaluated.

Criterion 3

Based on NUREG-1493, Regulatory Guide 1.163, and the rule

posting in the Federal Register (60 FR 49495), the margin for safety

presently provided is not significantly reduced by the proposed

change to a performance-based test interval for Types A, B, and C

tests. Although the changes allow more flexibility in scheduling

tests, the proposed amendment continues to ensure reactor

containment system reliability by periodic testing in full

compliance with 10 CFR 50, Appendix J Option B. Based on these

considerations, it is concluded that the changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

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

NRC Project Director: John N. Hannon.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota.

Date of amendment request: August 15, 1995, as supplemented

November 14, 1995.

Description of amendment request: The proposed amendment would

modify the Monticello Technical Specifications (TS) to: (1) revise the

main steam line isolation valve leak rate test acceptance criterion to

be based upon the combined maximum flow path leakage for all four main

steam lines of 46 standard cubic feet per hour (scfh) in lieu of the

current limit of 11.5 scfh per valve; (2) revise the operability test

interval for the drywell spray header and nozzles from 5 years to 10

years; and (3) revise TS 3/4.7.a.2, Primary Containment Integrity, to

remove information specific to the primary containment leakage rate

testing program and replace it with a commitment to abide by the

requirements of 10 CFR Part 50, Appendix J, Option B, Section III.A,

for Type A testing.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

a. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment is limited to changes to the surveillance

testing requirements applicable to the main steam line isolation

valves [MSIVs] allowable leakage criteria, drywell nozzles test

interval, and method of applying Appendix J test requirements. With

respect to monitoring main steam [line] isolation valve performance,

the proposed criteria are equivalent to the current criteria

ensuring that leakage past the valves would be within acceptable

limits under accident conditions. These surveillance tests are

performed while the plant is in a cold shutdown condition at a time

when the equipment is not required to be operable. Performance of

the tests themselves are not input or consideration in any accident

previously evaluated, thus the proposed change will not increase the

probability of any such accident occurring.

The proposed amendment will not adversely affect the function,

operation, or reliability of the equipment, nor will it diminish the

capability of the equipment to perform as required during an

accident.

[[Page 1633]]

Combining the maximum per valve leak rate into an overall maximum

leakage limit does not increase the overall permissible leakage and

thus has no significant impact on the consequences of previously

analyzed accidents since the combined leak rate of the main steam

line isolation valves, and thus the contribution of the valves to

overall primary containment leakage as used for analysis purposes,

is unchanged. Extending the drywell nozzle test interval has been

shown by industry experience to not compromise safety, and removing

the specifics of primary containment leakage testing from the

Technical Specifications and referencing 10 CFR Part 50 Appendix J

does not alter either how actual testing is accomplished nor the

acceptance criteria. It has been shown that adopting longer test

intervals based on performance, maintains the safety objective for

containment integrity while at the same time reducing the burden on

licensees, and provides a greater level of worker safety than that

provided by the previous rule.

Therefore, there will be no increase in post accident off-site

or on-site radiation dose as a result of this amendment. The

proposed amendment requires compliance with the regulatory

requirements of 10 CFR Part 50, Appendix J Option B, Section III.A,

for Type A testing that has previously been reviewed by the NRC and

found to be acceptable. Therefore, the amendment will not increase

the consequences of any accident previously evaluated.

b. The proposed amendment will not create the possibility of a

new or different kind of accident from any accident previously

analyzed.

The proposed amendment does not involve any modification to

plant equipment or operating procedures, nor will it introduce any

new equipment failure modes that have not been previously

considered. The proposed amendment is limited to changes in

surveillance test frequencies of tests performed while the plant is

in cold shutdown when the associated equipment is not required to be

operable. We therefore conclude the proposed changes will not create

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

accident previously analyzed.

c. The proposed amendment will not involve a significant

reduction in the margin of safety.

Combining the allowable leak rate for the MSIV's from a per

valve limit to an overall limit does not change the total allowable

leakage and therefore post accident dose levels remain unchanged.

Extending the drywell nozzle surveillance test interval from 5 to 10

years has been shown by industry experience to be acceptable.

Extending the intervals between containment integrated leakage tests

as authorized by 10 CFR Part 50, Appendix J, Option B, does not

change the acceptance criteria nor how testing is accomplished.

Based on these considerations, we conclude the proposed

amendment will not involve a significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

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

NRC Project Director: John N. Hannon.

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

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

California.

Date of amendment requests: December 19, 1995.

Description of amendment requests: The proposed amendments would

revise the combined Technical Specifications (TS) for the Diablo Canyon

Power Plant Unit Nos. 1 and 2 to relocate Technical Specification (TS)

6.5, ``Review and Audit,'' 6.8, ``Procedures and Programs,'' Sections

6.8.1c., 6.8.1d., 6.8.2, and 6.8.3, in accordance with guidance in an

NRC letter dated October 25, 1993, from William T. Russell to the

chairpersons of industry owners groups and the Commission's Final

Policy Statement on TS Improvements for Nuclear Power Reactors on

relocation of TS that do not satisfy the retention criteria. As part of

the relocation of TS 6.8.2, TS 6.1.1 would be revised to require that

proposed tests, experiments, or modifications that affect nuclear

safety be approved by the plant manager or his designee prior to

implementation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes simplify the Technical Specifications (TS),

meet regulatory requirements for relocated TS, and implement the

recommendations of: (1) the NRC's letter dated October 25, 1993,

from William T. Russell to the chairpersons of the industry owners

groups; (2) the Commissions's Final Policy Statement on TS

Improvements; and (3) the recently revised 10 CFR 50.36. Future

changes to these requirements will be controlled by 10 CFR 50.54 and

10 CFR 50.59. Any changes that reduce the effectiveness of the

Quality Assurance Program will be approved by the NRC prior to

implementation. The proposed changes are administrative in nature

and do not involve any modifications to any plant equipment or

affect plant operation.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes are administrative in nature, do not

involve any physical alterations to any plant equipment, and cause

no change in the method by which any safety-related system performs

its function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not alter the basic regulatory

requirements and do not affect any safety analyses. 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 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: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and

Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Project Director: William H. Bateman.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York.

Date of amendment request: September 15, 1995.

Description of amendment request: The licensee proposes to extend

the surveillance test intervals for the auxiliary electrical systems to

support 24-month operating cycles.

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:

Operation of the James A. Fitzpatrick plant in accordance with

the proposed

[[Page 1634]]

Amendment would not involve a significant hazards consideration as

defined in 10 CFR 50.92, since it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes increase the interval between auxiliary

electrical system functional tests and also propose additional

requirements for battery performance testing. These changes are

consistent with the guidance provided in Generic Letter 91-04. These

changes do not involve any special changes to the plant, nor do they

alter the way the auxiliary electrical system functions. Past

equipment performance indicates that the test acceptance criteria

has been consistently met, providing additional assurance that the

longer surveillance interval will not degrade system performance.

The proposed changes revise Bases section 4.9 to clarify battery

testing requirements and indicate consistence with the length of the

24 month operating cycle. Therefore, the proposed changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes increase the interval between auxiliary

electrical system functional tests and also propose additional

requirements for battery performance testing. These changes are

consistent with the guidance provided in Generic Letter 91-04. The

proposed changes do not change the ability of the auxiliary

electrical systems to provide electrical power during a design basis

accident. Past equipment performance indicates that the test

acceptance criteria has been consistently met, providing additional

assurance performance. The proposed changes do not modify the design

or operation of plant equipment, therefore, no new or different

failure modes are introduced. The proposed changes revise Basis

section 4.9 to clarify battery testing requirements and indicate

consistency with the length of the 24 month operating cycle.

Therefore, the proposed changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

The proposed changes increase the interval between auxiliary

electrical system functional tests and also propose additional

requirements for battery performance testing. These changes are

consistent with the guidance provided in Generic Letter 91-09. The

proposed changes do not alter the configuration of the auxiliary

electrical system nor change the manner in which the system

functions. Operation of the facility remains unchanged by the

proposed changes. An evaluation of past equipment performance

indicates that auxiliary electrical system operability is not time

dependent. The proposed changes revise Bases section 4.9 clarify

battery testing requirements and indicate consistency with the

length of the 24 month operating cycle. Therefore, a longer

surveillance test interval for the station batteries and LPCI [low-

pressure coolant injection] batteries will not degrade performance

of the auxiliary electrical system and will not involve a

significant reduction in a margin of safety.

The NRC staff has revised 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 request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Ledyard B. Marsh.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York.

Date of amendment request: October 25, 1995.

Description of amendment request: The licensee proposes to extend

the surveillance test intervals for the containment systems to support

24-month operating cycles.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 40.19(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92, since it would not:

1. Involve a significant increase in the probability of

consequences of an accident previously evaluated.

The proposed changes do not involve any physical changes to the

plant, do not alter the way the containment systems function, and

will not degrade the performance of the containment systems. The

type of testing and the corrective actions required if the subject

surveillance fail remains the same. The proposed changes do not

adversely affect the availability of the containment systems or

affect the ability of the systems to meet their design objectives. A

historical review of surveillance test results indicated that there

was no evidence of any failures which would invalidate the above

conclusions.

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

from any accident previously evaluated.

The proposed changes do not modify the design or operation of

the plant and therefore no new failure modes are introduced. No

changes are proposed to the type and method of testing performed,

only to the length of the surveillance interval. Past equipment

performance and on-line testing indicate that longer test intervals

will not degrade the containment systems. A historical review of

surveillance test results indicated that there was no evidence of

any failure which would invalidate the above conclusions.

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

Although the proposed changes will result in an increase in the

interval between surveillance tests, the impact on system

reliability is minimal. This is based on more frequent on-line

testing and the redundant design of the containment systems. A

review of past surveillance history has shown no evidence of failure

which would significantly impact the reliability of the containment

systems. Operation of the plant remains unchanged by the proposed

containment system surveillance test interval extensions. The

assumptions in the Plant Licensing Basis are not impacted. Therefore

the proposed changes do not result in a significant reduction in the

margin of safety.

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

this review, it appears that the three standards of 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: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Ledyard B. Marsh.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York.

Date of amendment request: November 30, 1995.

Description of amendment request: The licensee proposes to extend

the surveillance test intervals for the standby liquid control (SLC)

system to support 24 month operating cycles.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.19(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92 since it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

[[Page 1635]]

The proposed changes do not involve any physical changes to the

plant, do not alter any SLC system functions, and will not degrade

the performance of the SLC system. The type of testing and the

corrective actions required if the subject SLC surveillances fail

remain the same. The proposed changes do not adversely affect the

availability of the SLC system or the ability of the system to bring

the reactor from full power to a cold shutdown condition in the

unlikely event that control rods cannot be inserted. A historical

review of SLC surveillance test results indicated that there was no

evidence of any failures that would invalidate the above

conclusions.

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

from any accident previously evaluated.

The proposed changes do not introduce any failure mechanisms of

a different type than those previously evaluated since there are no

physical changes being made to the facility. No changes are proposed

to the type and method of testing performed, only to the length of

the surveillance interval. Past equipment performance and on-line

testing indicate the longer test intervals will not degrade SLC

equipment. A historical review of surveillance test results

indicated that there was no evidence of any failures that would

invalidate the above conclusions.

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

Although the proposed changes will result in an increase in the

interval between surveillance tests, the impact on system

reliability is minimal. This is based on more frequent on-line

testing of major system components and the redundant design of the

SLC system. A review of past SLC surveillance history has shown no

evidence of failures that would significantly impact the reliability

of the SLC system. The longer testing intervals do not significantly

impact the SLC safety margins for SLC normal operation, operation

with inoperable components, or sodium pentaborate solution as

described in the bases of the Technical Specifications. Operation of

the plant remains unchanged by the proposed SLC surveillance

interval extensions. The assumptions in the Plant Licensing Basis

are not impacted. Therefore, the proposed changes do not result in a

significant reduction in the margin of safety.

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

this review, it appears that the three standards of 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: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Ledyard B. Marsh.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York.

Date of amendment request: December 14, 1995.

Description of amendment request: The licensee proposes to

incorporate the inservice testing (IST) requirements of Section XI of

the American Society of Mechanical Engineers Boiler and Pressure Vessel

Code (ASME Code). The proposed change adds a new surveillance

requirement, 4.0.E, which refers to the requirements of Section XI of

the ASME Code and Addenda established by 10 CFR 50.55a(f). Ancillary

changes are also required since the proposed specification 4.0.E

replaces the surveillance testing requirements of safety related pump

and motor-operated valves and extends the surveillance testing

frequency of other components from once every month, to coincide with

the ASME Code Section XI requirements.

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:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92, since it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The changes identified in this proposed amendment revise

surveillance testing for various systems based upon the Section XI

of the American Society of Mechanical Engineers [***] Boiler and

Pressure Vessel [***] Code [ASME Code]. None of these changes

involves a hardware modification to the plant, a change to system

operation, a change to the manner in which the system is used, or a

change in the ability of the system to perform its intended

function.

The use of Section XI of the ASME [***] Code as a basis for

establishing surveillance testing and acceptance criteria will not

alter existing accident analyses. This has been acknowledged and

accepted by the NRC in the Standard Technical Specifications. The

change to surveillance testing frequencies reduces testing at power,

increases the availability of systems important to the mitigation of

a DBA [design-basis accident], and minimizes component degradation

due to excessive testing. The ASME [***] Code, Section XI testing

tracks component performance allowing identification of component

degradation and the code specifies that if a pump parameter enters

the alert range, then the testing frequency is doubled until the

cause of the degradation is determined and the condition corrected.

Similarly, if a valve stroke time degrades, the valve testing

frequency is increased to once per month until the cause is

determined and the condition corrected.

The editorial changes are strictly non technical in nature with

no effect on existing analyses. They clarify the Technical

Specifications by improving the legibility of this document.

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

from those previously evaluated.

The proposed changes involve no hardware changes, no changes to

the operation of the systems, and do not change the ability of the

systems to perform their intended functions. The use of ASME Section

XI as the basis for testing involves the same testing alignments and

practices previously used as part of either the IST program or

Technical Specification Surveillance Requirements. The editorial

changes have no effect on plant practices.

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

There are no hardware modifications, changes to system

operations, or effect on the ability of systems to perform their

intended function associated with the proposed changes. The proposed

changes to reference pump and valve testing to Section XI of the

ASME [***] Code and remove individual Surveillance Requirements in

the Technical Specifications does not relax any controls or

limitations. The resulting reduction in test frequency, while

reducing the possibility of detecting a degraded component prior to

failure, is offset by the increased availability of systems

important to plant safety and an associated reduction in component

wear and degradation due to excessive testing. Additionally, the

ASME testing program evaluates components for degraded performance

and will identify such degradation early. There are no safety

margins associated with the editorial corrections.

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 request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Ledyard B. Marsh.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public

Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,

Unit No. 1, Fairfield County, South Carolina.

Date of amendment request: December 8, 1995.

Description of amendment request: The proposed changes add a new

[[Page 1636]]

surveillance requirement to Technical Specification (TS) Section

4.1.2.2 and deletes TS Sections 3/4.1.2.3 and 3/4.1.2.4 associated with

the Borations Systems section. TS Section 3/4.9.3 is being revised to

assure only one charging pump is capable of Reactor Coolant System

injection in the applicable modes and to add a new surveillance

requirement to demonstrate this assurance. TS Section 4.5.2.f is being

revised to delete specific Emergency Core Cooling System pump testing

acceptance criteria and reference acceptance criteria located in the

plant Inservice Testing Program. In addition, the licensee has proposed

changes to the bases.

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 probability or consequences of an accident previously

evaluated is not significantly increased.

The implementation of the above described TS changes will have

no impact on the probability of an accident occurring. The testing

of the ECCS pumps at a more appropriate point on their

characteristic curve is not a precursor to an accident. There is no

hardware, software, or testing methodology change proposed that

would decrease confidence in the reliability of these systems/

components.

The proposed revision to the ECCS Pump testing surveillance will

allow greater flexibility for testing and will provide more useful

information about the performance capabilities of those pumps.

The deletion of the Reactivity Control System Specifications

(Charging Pumps - Operating and Charging Pumps - Shutdown) will have

no impact on the capability of the Charging/SI pumps to perform

their design function. The additional Action Statement and

Surveillance for low temperature overpressure (LTOP) assure that

safety analyses remain valid and initial conditions are not changed.

The additional Surveillance Requirement for Boration Systems assures

that one charging pump will be operable during Modes 5 and 6.

2. The proposed license amendment does not create the

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

previously evaluated.

This proposed TS change does not involve any changes to station

hardware, software, or operating practices. The changes do provide

for a revision to the testing methodology used in demonstrating the

capability of the ECCS pumps.

This methodology will test the ECCS pumps at a point on the

pump's characteristic curve that will more reliably indicate the

pump's continued operability at or near the parameters the pump

would be required to provide during a postulated accident.

The deletion of the Reactivity Control System Specifications

(Charging Pumps - Operating and Charging Pump - Shutdown) will not

provide additional challenges to the capability of the plant to meet

normal operational needs or mitigate the conditions of a design

basis accident. The ECCS Subsystems TS provide similar surveillance

requirements to insure continued operability of the Charging/SI

pumps. The LTOP TS will now provide requirements to assure that

design assumptions are not challenged and RCS integrity is

maintained.

Therefore, as the above described change has no impact on plant

performance, the possibility of a new or different kind of accident

being created as a result of this change is negligible.

3. The proposed license amendment does not involve a significant

reduction in a margin of safety.

The change in testing philosophy for ECCS pumps should bring an

increase in margin of safety, since testing will be conducted at

reference flow points closer to actual pump parameters for accident

conditions. For the Residual Heat Removal Pumps this will be

conducted quarterly and for the centrifugal charging pumps, they

will be tested quarterly on minimum flow and each refueling outage

at substantial flow per the Inservice Testing Program.

The surveillance requirements of TS 3/4.1.2.3 and TS 3/4.1.2.4

are essentially the same as those in 3./4.5.2 and 3/4.5.3 (ECCS

Subsystems), and the deletion of these requirements will have no

adverse impact on margin on safety. The addition of the Action

Statement and Surveillance Requirements to 3/4.4.9.3 (Overpressure

Protective Systems) provide additional requirements to supplement

those above to assure RCS integrity is maintained for all

operational modes. The addition of the Surveillance Requirement to

3/4.1.2.1 will provide assurance that reactivity control can be

maintained for Modes 5 and 6 through the charging system flow path.

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: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Project Director: Frederick J. Hebdon.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama.

Date of amendments request: December 19, 1995.

Description of amendments request: The proposed amendments would

replace the requirements associated with the Control Room Emergency

Ventilation System with requirements related to the operation of the

Control Room Emergency Filtration/Pressurization System and Control

Room Air Conditioning System. These changes are technically consistent

with the requirements of NUREG-1431, Revision 1, ``Westinghouse

Standard Technical Specifications,'' issued on April 7, 1995. Also, a

one-time extension to the allowable outage time for the control room

recirculation filtration system is included to facilitate

implementation of design modifications to enhance the reliability of

the control room air conditioning system during the spring of 1996.

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:

Based on the preceding evaluation, the following conclusions are

provided with respect to the criteria contained in 10 CFR 50.92.

(1) The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated in

the FSAR [Final Safety Analysis Report]. The proposed changes have

no impact on the probability of an accident. The control room

ventilation systems are support systems which have a role in the

detection and mitigation of accidents but do not contribute to the

initiation of any accident previously evaluated. Reorganizing the

technical specifications by functions have no impact on the course

of any accidents previously evaluated. The other changes which are

being made improve the ability to mitigate fuel handling accidents.

Specifying an allowed outage time (AOT) of 30 days for the cooling

of recirculated air while one train is inoperable is based on the

significance of the cooling function but does represent an increase

in the allowed outage time and thus an increase in the probability

that the functions could be unavailable. This increase is not

considered significant based on several factors including: the

design is based on the worst postulated meteorological conditions;

generally, less than design cooling is required and a partial

failure in the system may have no impact; and unavailability failure

does not create an immediate irreversible impact (i.e., temperature

will increase slowly over a period of time); the system could be

restored or its loss mitigated without any impact on the course or

whatever accident is being considered; and the extended AOT would

allow more opportunity to perform major required maintenance and

thus may provide an overall improvement in equipment reliability.

In addition, the one-time change to the AOT for the

recirculation filtration will not

[[Page 1637]]

significantly increase the probability or consequences of an accident

due to the low probability of an event result[ing] in an airborne

release of radioactivity. Such an event requires multiple failures

of safety systems that are governed by technical specifications not

affected by these changes. In addition, compensatory measures have

been identified that limit the potential exposure of control room

operators in response to a postulated release.

The net effect of these changes is not significant and, as a

result, the changes do not involve a significant increase in the

consequences of an accident previously evaluated.

(2) The proposed changes to the Technical Specifications do not

increase the possibility of a new or different kind of accident than

any accident already evaluated in the FSAR. No new limiting single

failure or accident scenarios have been created or identified due to

the proposed changes. Safety-related systems are expected to perform

as designed. Although the changes could have a minor impact on the

air conditioning system availability, the changes do not create the

possibility of a new or different kind of accident from any

previously evaluated.

(3) The proposed changes do not involve a significant reduction

in the margin of safety. The changes proposed do not alter the

environmental conditions which are to be maintained in the control

room during normal operations and following an accident. As a

result, the margin of safety for these functions remains the same.

Although there is a potential impact on the air conditioning

system's postulated availability, there is no impact on the accident

analyses. Further, although the one-time AOT extension for the

recirculation filtration system increases the system unavailability

during the planned CRACS [Control Room Air Conditioning System]

design changes, the net effect is a benefit to plant safety due to

the enhancement to control room cooling capability. Thus, even if

system availability issues were considered an aspect of margin of

safety, the proposed 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 proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201.

NRC Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,

Alabama.

Date of amendment request: December 8, 1995 (TS 364).

Description of amendment request: The licensee proposes revision of

Units 1, 2, and 3 Technical Specifications (TS) Section 4.7.A to

implement the revision to 10 CFR 50, Appendix J. The new rule (Option

B) provides a voluntary performance-based testing option for

containment leak rate testing. Option B containment leak rate testing

requirements are based on system and component performance in lieu of

compliance with the current prescriptive requirements.

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:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment to TS Section 4.7.A is in accordance with

Option B to 10 CFR 50, Appendix J. The proposed amendment adds a

voluntary performance based option for containment leak rate

testing. The changes being proposed do not affect the precursor for

any accident or transient analyzed in Chapter 14 of the BFN [Browns

Ferry Nuclear Plant] Updated Final Safety Analysis Report (UFSAR).

The proposed change does not increase the total allowable primary

containment leakage rate. The proposed change does not reflect a

revision to the physical design and/or operation of the plant.

Therefore, operation of the facility in accordance with the proposed

change does not affect the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed amendment to TS Section 4.7.A is in accordance with

the new performance-based option (Option B) to 10 CFR 50, Appendix

J. The changes being proposed will not change the physical plant or

the modes of operation defined in the facility license. The proposed

changes do not increase the total allowable primary containment

leakage rate. The changes do not involve the addition or

modification of equipment, nor do they alter the design or operation

of plant systems. Therefore, operation of the facility in accordance

with the proposed change does not create the possibility of a new or

different kind of accident from any previously evaluated.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change to TS Section 4.7.A is in accordance with

the new option to 10 CFR 50, Appendix J. The proposed option is

formulated to adopt performance-based approaches. This option

removes the current prescriptive details from the TS. The proposed

changes do not affect plant safety analyses or change the physical

design or operation of the plant. The proposed change does not

increase the total allowable primary containment leakage rate.

Therefore, operation of the facility in accordance with the proposed

change does not involve a significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear

Power Station, Unit No. 1, Ottawa County, Ohio.

Date of amendment request: December 12, 1995.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.6.1.1, Containment Systems--

Primary Containment--Containment Integrity; TS 3/4.6.1.2, Containment

Systems--Containment Leakage; TS 3/4.6.1.6, Containment Systems--

Containment Vessel Structural Integrity; TS 3/4.6.5.3, Containment

Systems--Shield Building Structural Integrity; and associated Bases.

The proposed revisions adopt the provisions of Appendix J, Option B for

Type A containment leakage testing as modified by approved exemptions

and in accordance with the guidance of Regulatory Guide 1.163. The

licensee proposes to delete surveillance requirement (SR) 4.6.1.2, SR

4.6.1.2.b, SR 4.6.1.2.c, and SR 4.6.1.2.i since these requirements

contain details that are now included in standards that are referenced

by Regulatory Guide 1.163. TS 3/4.6.1.6 and TS 3/4.6.5.3 which address

containment building and shield building structural integrity are

proposed to be deleted since the requirements are addressed in revised

TS 3.6.1.2.a. The licensee proposes to delete the exemption included in

Bases

[[Page 1638]]

3/4.6.1.2 since it is no longer applicable. Additionally, the licensee

proposes to modify the Action statement associated with TS 3.6.1.2 to

reflect the action to take if the as-left rather than the as-found

leakage exceeds 0.75 La.

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:

Toledo Edison has reviewed the proposed changes and determined

that a significant hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in

accordance with the changes would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because accident initiators,

conditions, or assumptions are not affected by the proposed changes.

The proposed changes to the Technical Specifications implement

10 CFR 50 Appendix J Option B for Type A testing, including visual

examinations of the containment vessel and shield building, and make

various administrative changes to the Technical Specifications and

associated Technical Specification Bases. Therefore, as stated

above, these proposed changes do not affect accident initiators,

conditions, or assumptions.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because the proposed changes do not

change the source term, containment isolation, or allowable

releases.

The proposed changes involve containment leakage testing and

test frequency. The allowable containment leakage rates presently

specified in the Technical Specifications remain unchanged.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because no new

accident initiators or assumptions are introduced by the proposed

changes.

3. Not involve a significant reduction in a margin of safety,

for the reasons cited below.

The proposed changes involve containment leakage testing and

test frequency. The allowable containment leakage rates presently

specified in the Technical Specifications remain unchanged. The

Technical Specifications, under the proposed changes, will continue

to ensure containment system reliability by periodic testing

performed in full compliance with 10 CFR 50 Appendix J.

As stated in the Federal Register publication of the final rule,

60 FR 49495 dated September 26, 1995, the final rule improves the

focus of the regulations by eliminating prescriptive requirements

that are marginal to safety. Further, the final rule allows test

intervals to be based on system and component performance and

provides licensees greater flexibility for cost-effective

implementation methods of regulatory safety objectives. The final

rule publication also discusses the following specific findings

documented in NUREG-1493, ``Performance-Based Containment Leak-Test

Program,'' September, 1995, which justify the proposed change in

frequency of Type A Integrated Leak Rate Testing (ILRT):

1. The fraction of leakages detected only by ILRT's is small, on

the order of a few percent.

2. Reducing the frequency of ILRT testing from 3 every 10 years

to one every 10 years leads to a marginal increase in risk.

3. At a frequency of one test every 10 years, industry-wide

occupational exposure would be reduced by 0.087 person-sievert (8.7

person-rem) per year.

Based on these considerations, it is concluded that the proposed

changes do not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606.

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

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

NRC Project Director: Gail H. Marcus.

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

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

Manitowoc County, Wisconsin.

Date of amendment request: December 13, 1995.

Description of amendment request: The proposed amendments will

modify Technical Specification (TS) Sections 15.1, ``Definitions,''

15.2, ``Safety Limits and Limiting Safety System Settings,'' 15.3,

``Limiting Conditions for Operation,'' and 15.6, ``Administrative

Controls.'' The proposed changes would modify the TSs to account for

the creation and maintenance of a Core Operating Limits Report.

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 this facility under the proposed Technical

Specifications will not create a significant increase in the

probability or consequences of an accident previously evaluated.

The relocation of the cycle-specific parameters from the Point

Beach Nuclear Plant (PBNP) Technical Specifications to the Core

Operating Limits Report (COLR) has no impact on plant operation or

accident analyses. The proposed changes are administrative in

nature. The Technical Specifications will continue to require

operation within the core operational limits for each cycle reload

calculated by the NRC-approved reload design methodologies. The

appropriate actions required if limits are exceeded will remain in

the Technical Specifications. The reload report presents the results

of a cycle-specific evaluation of accidents and transients addressed

in the PBNP Final Safety Analysis Report (FSAR). The cycle-specific

evaluation demonstrates that changes in the unit's fuel cycle design

and corresponding COLR parameters do not involve a significant

increase in the probability or consequences of an accident

previously evaluated. Therefore, these changes do not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

2. Operation of this facility under the proposed Technical

Specifications will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed change to relocate the cycle-specific parameters

from the Technical Specifications to the COLR is administrative in

nature. No change to the design, configuration, or method of

operation of the plant is made by this change. The cycle-specific

parameters will be determined using NRC-approved methodologies. The

Technical Specifications will continue to require operation within

the core operating limits and appropriate actions will be taken if

the limits are exceeded.

Therefore, these changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

3. Operation of this facility under the proposed Technical

Specifications will not create a significant reduction in a margin

of safety.

Existing Technical Specification operability and surveillance

requirements are not reduced by the proposed changes to relocate

cycle-specific parameters from the Technical Specifications to the

COLR. The cycle-specific COLR limits for reloads will continue to be

developed based on NRC-approved methodologies, thereby maintaining

accepted margins of safety. The Technical Specifications will still

require that the core be operated within these limits and specify

appropriate actions to be taken if the limits are violated. Each

reload undergoes a 10 CFR 50.59 safety review to assure that

operating the unit within the cycle-specific limits will not involve

a significant reduction in a margin of safety. 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 proposes to determine that the

amendment request involves no significant hazards consideration.

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

[[Page 1639]]

Sixteenth Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,

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

NRC Project Director: Gail H. Marcus.

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

Generating Station, Coffey County, Kansas.

Date of amendment request: December 13, 1995.

Description of amendment request: This license amendment request

proposes to revise the 125-volt D.C. Sources Technical Specifications

(3.8.2.1 and 3.8.2.2) to include provisions for installed spare

chargers, which will be added to the plant design during the next

refueling outage. The Onsite Power Distribution Technical

Specifications 3.8.3.1 and 3.8.3.2 would be revised to indicate that

spare chargers may be connected in place of the primary chargers.

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.

These proposed technical specification changes do not alter the

plant design bases nor do they involve any hardware changes that

significantly increase the probability of any event initiators.

There will be no change to normal plant operating parameters or

accident mitigation capabilities. There will be no increase in the

consequences of any accident or equipment malfunction.

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

or different kind of accident from any accident previously

evaluated.

The proposed technical specification changes do not involve any

design bases changes nor are there any changes to the method by

which any safety-related plant system performs its safety function.

The normal manner of plant operation is unaffected. No new accident

scenarios, transient precursors, failure mechanisms, or limiting

single failures are introduced as a result of these changes.

3. The proposed change does not involve a significant reduction

in a margin of safety.

There will be no effect on the manner in which safety limits or

limiting safety system settings are determined, nor will there be

any effect in those plant systems necessary to assure the

accomplishment of protection functions. There will be no impact on

DNBR [departure from nucleate boiling ratio] limits, FQ, F-

delta-H, LOCA [loss-of-coolant accident] PCT [peak cladding

temperature], peak local power density or any other 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 locations: Emporia State University,

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

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

66621.

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

Trowbridge, 2300 N Street, NW., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

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

Generating Station, Coffey County, Kansas.

Date of amendment request: December 13, 1995.

Description of amendment request: This change request proposes

revising the minimum and maximum flow requirements for the centrifugal

charging pumps (CCPs) and safety injection pumps (SIPs) specified in

Technical Specification Surveillance Requirement 4.5.2.h. Specifically,

the proposed changes would:

(1) Decrease the minimum limits on the sum of the injection line

flow rates, excluding the highest flow rate, from 346 gpm to 330 gpm

for the CCPs and from 459 gpm to 450 gpm for the SIPs.

(2) Revise the maximum pump flow rate for the SIP from 665 to 670

gpm, but retain the CCPs maximum pump flow rate at its current value of

556 gpm.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change will not result in a condition where the

material or construction standards applicable prior to the change

are altered. The ECCS [emergency core cooling system] system

integrity is not affected by this change, and this change will not

affect the ability of the ECCS to fulfill its design functions. This

change will modify the pump surveillance criteria to prevent pump

runout during the test, but will not affect the method of operation

of the system and will not alter the testing method for the pumps.

This change will slightly alter the acceptance criteria of the test,

but the changes have been determined to be enveloped by the ECCS

pump flow and balance criteria assumed in the safety analyses

described in the USAR [Updated Safety Analysis Report]. This change

will not affect the ability of the ECCS to mitigate the consequences

of any previously evaluated accident. The proposed change will not

alter, degrade or prevent the response of the ECCS to any accident

scenarios evaluated in the USAR. Therefore, neither the probability

of occurrence nor the consequences of any accident previously

evaluated in the USAR will be increased by this change.

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

or different kind of accident from any accident previously

evaluated.

The proposed change will alter the existing ECCS pump flow test

to prevent pump runout during the test by slightly altering the

acceptance criteria of the test. However, the proposed changes have

been determined to be enveloped by the ECCS pump flow and balance

criteria assumed in the safety analyses described in the USAR. This

change will not create a new type of accident or malfunction, and

the method and manner of plant operation remains unchanged. This

change will not alter the safety functions of the ECCS. The safety

design bases in the USAR have not been altered, and no new or

different accident scenarios, transient precursors, failure

mechanisms, or limiting single failures will be introduced as a

result of this change. Therefore, the possibility of a new or

different kind of accident other than those already evaluated will

not be created by this change.

3. The proposed change does not involve a significant reduction

in a margin of safety.

There are no changes being made to any safety limits or safety

system settings that would adversely impact plant safety. This

proposed change will have no affect on the availability, operability

or performance of any safety-related system or component. The

analysis results and conclusions of the accidents presented in the

current USAR would not be adversely affected by the revised

surveillance requirements for the ECCS. This conclusion is drawn

based on the evaluation that confirms that the actual ECCS flow

characteristics remain consistent with assumptions used in the WCGS

[Wolf Creek Generating Station] accident analyses. Specifically, the

accident analyses which are limiting with minimized ECCS flow have

already been analyzed using revised ECCS flows that were developed

based on a more conservative minimum flow than the proposed minimum

ECCS flow requirement. For the analyses which are limiting with a

higher ECCS flow, the evaluation indicated that a higher pump runout

limit proposed for the SIPs would have insignificant effect on the

results and conclusions of the analyses. The evaluation also

indicated that the ECCS pump operability would not be a concern as a

result of increasing the SIPs runout limit because the available

runout margin is sufficient to accommodate the cumulative effect of

the ECCS performance issues. Based on these reasons, it is concluded

that

[[Page 1640]]

implementation of the proposed changes will have no adverse impact on

the ECCS subsystems' operability and their intended safety function.

Therefore, the proposed change would not result in 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 locations: Emporia State University,

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

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

66621.

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

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

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

Generating Station, Coffey County, Kansas.

Date of amendment request: December 13, 1995.

Description of amendment request: This license amendment request

proposes revising Surveillance Requirement 4.1.3.1.3 to delete the

requirement for performing the control rod drop surveillance test with

Tavg greater than or equal to 551 deg.F. This would allow

performing this test with Tavg below 551 deg.F. This change will

also add justification for performing the rod drop test with Tavg

below 551 deg.F to Bases Section 3/4.1.3, ``Movable Control

Assemblies.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change will not result in a condition where the

material or construction standards applicable prior to the change

are altered. The rod control system integrity is not affected by

this change, and this change will not affect the ability of the

system to fulfill its design function. This change will allow the

control rod drop test to be performed at lower temperatures than

currently allowed, but will not affect the method of operation of

the system and will not alter the drop time criterion of the test.

This change will not affect any fission product barrier, and will

not affect the integrity of any fuel assembly or the reactor

internals. Thus this change will not affect the ability of the rod

control system to mitigate the consequences of any previously

evaluated accident. The proposed change will not alter, degrade or

prevent the response of the rod control system to any accident

scenarios evaluated in the USAR [Updated Safety Analysis Report].

Therefore, neither the probability of occurrence nor the

consequences of any accident previously evaluated in the USAR will

be increased by this change.

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

or different kind of accident from any accident previously

evaluated.

The proposed change will alter the existing rod drop test to

allow the test to be performed over a range of temperatures, but

will not alter the rod drop time criterion of the test. This change

will not create a new type of accident or malfunction, and the

method and manner of plant operation remains unchanged. This change

will not alter the safety functions of the rod control system. The

safety design bases in the USAR have not been altered, and no new or

different accident scenarios, transient precursors, failure

mechanisms, or limiting single failures will be introduced as a

result of this change. Therefore, the possibility of a new or

different kind of accident other than those already evaluated will

not be created by this change.

3. The proposed change does not involve a significant reduction

in a margin of safety.

There are no changes being made to any safety limits or safety

system settings that would adversely impact plant safety. This

proposed change will have no affect on the availability, operability

or performance of any safety-related system or component. The change

will not prevent inspections or surveillances required by the

technical specifications, and does not alter the rod drop time

criterion specified in the technical specifications. Performance of

the rod drop tests at other temperatures allows an alternative

method to verify that the rod drop time currently specified in the

technical specifications and used in the safety analyses continues

to be valid. Therefore, the proposed change would not result in 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 locations: Emporia State University,

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

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

66621.

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

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

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.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic City Electric Company, Docket

Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2

and 3, York County, Pennsylvania.

Date of amendment request: November 21, 1995

Brief description of amendment request: The proposed amendments

would revise surveillance requirements for the high pressure coolant

injection and reactor core isolation cooling systems and would make an

administrative change to Section 5.5.7 of the technical specifications

to eliminate reference to a section which was previously eliminated.

Date of publication of individual notice in Federal Register:

December 5, 1995 (60 FR 62271).

Expiration date of individual notice: January 3, 1996.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (Regional Depository) Education

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

Pennsylvania 17105.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic City Electric Company, Docket

Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2

and 3, York County, Pennsylvania.

Date of amendment request: November 30, 1995.

[[Page 1641]]

Brief description of amendment request: The proposed amendments

would revise the minimum allowable control rod scram accumulator

pressure and charging water header pressure from a value of 955 psig to

a value of 940 psig.

Date of publication of individual notice in Federal Register:

December 8, 1995 (60 FR 63073).

Expiration date of individual notice: January 8, 1996.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (Regional Depository) Education

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

Pennsylvania 17105.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic City Electric Company, Docket

Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2

and 3, York County, Pennsylvania.

Date of amendment request: December 19, 1995.

Brief description of amendment request: The proposed amendment

would revise the ventilation filter test program (VFTP) bypass and

penetration leakage test acceptance criteria from less than 0.05

percent to less than 1.0 percent. The change corrects an administrative

error that occurred during the development of the Peach Bottom Improved

Technical Specifications which were issued as Amendments 210 and 214 to

the Peach Bottom licenses on August 30, 1995.

Date of publication of individual notice in Federal Register:

December 27, 1995 (60 FR 66997).

Expiration date of individual notice: January 25, 1996.

Local Public Document Room location: Government Publications

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

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

Pennsylvania 17105.

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 local public document rooms for

the particular facilities involved.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs

Nuclear Power Plant, Unit No. 1, Calvert County, Maryland.

Date of application for amendment: October 20, 1995.

Brief description of amendment: The one-time amendment revises the

Calvert Cliffs Nuclear Power Plant, Unit No. 1 Technical Specifications

by extending certain 18-month instrument surveillance intervals by a

maximum of 39 days to March 31, 1996. This amendment will be superseded

by Amendment No. 208 when it is implemented prior to restart from the

Unit No. 1 spring 1996 refueling outage.

Date of issuance: December 28, 1995.

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

30 days.

Amendment No.: 209.

Facility Operating License No. DPR-53: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 27, 1995 (60

FR 58396).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 28, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs

Nuclear Power Plant, Unit No. 2, Calvert County, Maryland.

Date of application for amendment: October 2, 1995.

Brief description of amendment: The amendment revises the Technical

Specifications regarding allowable outage time (AOT) associated with

the control room emergency ventilation system. It extends the AOT for

one train from 7 days to 30 days on a one-time basis (for the loss of

the emergency power supply only) to allow for modifications during the

upcoming Unit No. 1 refueling outage in the spring of 1996.

Date of issuance: December 19, 1995.

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

the Unit No. 1 spring 1996 refueling outage.

Amendment No.: 187.

Facility Operating License No. DPR-69: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 8, 1995 (60 FR

56363).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 19, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units

1 and 2, Rock Island County, Illinois.

Date of application for amendments: September 10, 1993, as

supplemented on June 16, 1995.

Brief description of amendments: This application upgrades the

current custom Technical Specifications (TS) for Dresden and Quad

Cities to the Standard Technical Specifications contained in NUREG-

0123, ``Standard Technical Specification General Electric Plants BWR/

4.'' This application upgrades only Section 3/4.8 (Plant Systems).

Date of issuance: December 19, 1995.

[[Page 1642]]

Effective date: Immediately, to be implemented no later than June

30, 1996.

Amendment Nos.: 144, 138, 166, and 162.

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: July 19, 1995 (60 FR

37086).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 19, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021.

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

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

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units

1 and 2, Rock Island County, Illinois.

Date of application for amendments: September 15, 1995.

Brief description of amendments: The amendments upgrade the current

custom Technical Specifications (TS) for Dresden and Quad Cities to the

Standard Technical Specifications contained in NUREG-0123, ``Standard

Technical Specification General Electric Plants BWR/4.'' The

application dated September 15, 1995, contains some of the TSUP open

items from previous Dresden and Quad Cities TS amendments issued by the

NRC.

Date of issuance: December 19, 1995.

Effective date: Immediately, to be implemented no later than June

30, 1996.

Amendment Nos.: 145, 139, 167 and 163

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 5, 1995 (60 FR

52220).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 19, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021.

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

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

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units

1 and 2, Rock Island County, Illinois.

Date of application for amendments: September 17, 1993, as

supplemented July 28, 1995.

Brief description of amendments: This application upgrades the

current custom Technical Specifications (TS) for Dresden and Quad

Cities to the Standard Technical Specifications contained in NUREG-

0123, ``Standard Technical Specification General Electric Plants BWR/

4.'' This application upgrades only Section 3/4.5 (Emergency Core

Cooling Systems).

Date of issuance: December 27, 1995.

Effective date: Immediately, to be implemented no later than June

30, 1996.

Amendment Nos.: 146, 140, 168, and 164.

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 16, 1995 (60 FR

42599).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 27, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021.

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: November 14, 1995.

Brief description of amendments: These amendments change the

implementation dates of all previous TSUP amendments from December 31,

1995, to no later than June 30, 1996.

Date of issuance: December 29, 1995.

Effective date: December 29, 1995.

Amendment Nos.: 147 and 141.

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

revised the license.

Date of initial notice in Federal Register: November 29, 1995 (60

FR 61272).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 29, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station,

Unit 1, LaSalle County, Illinois.

Date of application for amendment: October 2, 1995.

Brief description of amendment: The amendment revises the safety/

relief valve (SRV) safety function lift setting allowable tolerance

band from -3/+1% to 3% and includes a requirement for the

lift settings to be within 1% of the technical

specification limit following testing.

Date of issuance: January 3, 1996.

Effective date: Upon date of issuance; shall be implemented prior

to the restart of Unit 1 from its seventh refueling outage.

Amendment No.: 108.

Facility Operating License No. NPF-11: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 27, 1995 (60

FR 58398).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 3, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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 application for amendments: September 5, 1995.

Brief description of amendments: In Section 5.2.5 of the Catawba

Safety Evaluation Report (SER, NUREG-0954), the NRC staff identified

that the air particulate monitors (EMF38, at both Units 1 and 2), are

designed to seismic Category I requirements. A recent engineering

review by the licensee determined that documentation did not exist to

show these monitors are designed to seismic Category I requirements. In

a submittal dated September 8, 1994, the licensee proposed a technical

justification for not requiring the subject monitors to be

[[Page 1643]]

seismic Category I, and by letter dated September 5, 1995, provided

additional justification and requested amendments to the licenses for

both Units 1 and 2. The NRC staff has reviewed the licensee's

justification and concludes that the containment air particulate

monitors at Catawba do not have to meet seismic Category I

requirements. The bases for this conclusion are included in the NRC

staff's Safety Evaluation.

Date of issuance: December 29, 1995.

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

30 days.

Amendment Nos.: Unit 1--140; Unit 2--134.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Updated Final Safety Analysis Report.

Date of initial notice in Federal Register: November 28, 1995 (60

FR 58690).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 29, 1995 and an Environmental

Assessment dated December 22, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730.

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

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

Date of application for amendments: September 1, 1995, as

supplemented by letters dated October 17 and November 15, 1995.

Brief description of amendments: The requested changes would revise

Technical Specification (TS) 6.9.1.9 to include references to updated

or recently approved methodologies used to calculate cycle-specific

limits contained in the Core Operating Limits Report (COLR). The

subject references have previously been reviewed and approved by the

NRC staff.

Date of issuance: December 19, 1995.

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

30 days.

Amendment Nos.: Unit 1--160; Unit 2--142.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 25, 1995 (60 FR

54718).

The October 17 and November 15, 1995, letters provided clarifying

information that did not change the scope of the September 1, 1995,

application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 19, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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

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

Date of application for amendments: January 12, 1995, as

supplemented by letter dated June 29, 1995.

Brief description of amendments: The amendments would revise and

clarify portions of Technical Specification Section 6.0,

``Administrative Controls.''

Date of issuance: December 19, 1995.

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

30 days.

Amendment Nos.: Unit 1--161; Unit 2--143.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: March 15, 1995 (60 FR

14018).

The June 29, 1995, letter provided clarifying information that did

not change the scope of the January 12, 1995, application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a

Safety Evaluation dated December 19, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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 application of amendments: July 26, 1995, as supplemented

by letter dated November 20, 1995.

Brief description of amendments: The amendments add a footnote to

Technical Specification 3.7.8 to provide for a one-time extension of

the allowable outage time from 72 hours to 7 days for the Oconee

overhead emergency power path to be inoperable, so that proposed

modifications to the degraded grid protection system and the external

grid trouble protection system may be performed.

Date of Issuance: December 27, 1995.

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

30 days from the date of issuance.

Amendment Nos.: Unit 1--213; Unit 2--213; Unit 3--210.

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 16, 1995 (60 FR

42601).

The November 20, 1995, letter provided clarifying information that

did not change the scope of the July 26, 1995, application and the

proposed no significant hazards consideration determination. The

Commission's related evaluation of the amendments is contained in a

Safety Evaluation dated December 27, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691.

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

No. 2, Pope County, Arkansas.

Date of application for amendment: July 19, 1995.

Brief description of amendment: The amendment reduced the

requirements associated with the exercise frequency of control element

assemblies from once per 31 days to once per 92 days.

Date of issuance: December 22, 1995.

Effective date: December 22, 1995, to be implemented within 30

days.

Amendment No.: 173.

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

Technical Specifications.

Date of initial notice in Federal Register: October 11, 1995 (60 FR

52929).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

No. 2, Pope County, Arkansas.

Date of application for amendment: April 4, 1995.

Brief description of amendment: The amendment revises surveillance

[[Page 1644]]

requirements associated with the main turbine steam valves.

Date of issuance: December 22, 1995.

Effective date: December 22, 1995, to be implemented within 30

days.

Amendment No.: 174.

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

Technical Specifications.

Date of initial notice in Federal Register: July 5, 1995 (60 FR

35069).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County, Florida.

Date of application for amendments: September 11, 1995, as

supplemented by letter dated November 22, 1995.

Brief description of amendments: These amendments revise the

emergency diesel generator testing requirements to incorporate the

recommendations of Generic Letters 93-05 and 94-01.

Date of issuance: December 28, 1995.

Effective date: December 28, 1995.

Amendment Nos. 181 and 175.

Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 11, 1995 (60 FR

52930).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 28, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and

50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,

Georgia.

Date of application for amendments: December 2, 1994.

Brief description of amendments: The amendments replace Appendix B,

``Environmental Technical Specifications,'' with an Environmental

Protection Plan (Nonradiological) and revise the Operating Licenses to

reflect these changes.

Date of issuance: December 19, 1995.

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

30 days.

Amendment Nos.: Unit 1--199; Unit 2--140.

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

revised the Technical Specifications and Operating Licenses.

Date of initial notice in Federal Register: January 4, 1995 (60 FR

502).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 19, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook,

Nuclear Plant, Unit No. 1, Berrien County, Michigan.

Date of application for amendment: April 13, 1995, as supplemented

August 28 and October 27, 1995.

Brief description of amendment: The amendment modifies the

Technical Specifications to allow use of laser-welded sleeves to repair

defective steam generator tubes.

Date of issuance: January 4, 1996.

Effective date: January 4, 1996, with full implementation within 45

days.

Amendment No.: 205.

Facility Operating License No. DPR-58. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 6, 1995 (60 FR

29877).

The August 28 and October 27, 1995, supplements provided clarifying

information and updated Technical Specification pages. These

supplements did not change the proposed no significant hazards

considerations determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 4, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085.

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

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

Date of application for amendment: August 31, 1995, as supplemented

December 5, 1995.

Brief description of amendment: The amendment modifies the

definition of HOT SHUTDOWN and COLD SHUTDOWN to specify that the

definitions are not applicable during the performance of an inservice

hydrostatic and leak test (IHLT). Technical Specification Section 3.6.B

and 4.6.B is modified by adding Section 3.6.B.1.b and 4.6.B.1.b to

identify the requirements that must be satisfied to consider the

reactor in COLD SHUTDOWN during the performance of an IHLT. In

addition, the amendment changes temperature specific requirements on

several pages to mode or condition specific requirements; makes several

editorial changes; and changes the associated Bases.

Date of issuance: December 29, 1995.

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

within 30 days.

Amendment No.: 90.

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

Technical Specifications.

Date of initial notice in Federal Register: September 27, 1995 (60

FR 49940).

The December 5, 1995, 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 December 29, 1995.

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, CT 06360.

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

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

Date of application for amendment: May 1, 1995.

Brief description of amendment: The amendment revises the Technical

Specifications to extend the interval for performance of selected

surveillances to accommodate a 24-month fuel cycle. Specifically, this

amendment changes the definition for a refueling interval, changes the

BASES for surveillances that are performed at least once each fuel

cycle and changes the surveillance frequencies for:

(1) The flow path tests of the boron injection system,

(2) The operability tests of the digital rod position indicatiors,

(3) The drop time of the full-length shutdown and control rods,

(4) The channel calibration of the loose-part detection system,

[[Page 1645]]

(5) The channel calibration of the seismic monitoring

instrumentation,

(6) The activation of the pumps and the flow path tests of the

valves in the containment quench and recirculation spray systems and

(7) The tests of the intended actuation positions of the

containment isolation valves.

Date of issuance: December 28, 1995.

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

within 90 days.

Amendment No.: 122.

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

Technical Specifications.

Date of initial notice in Federal Register: November 27, 1995 (60

FR 58402).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 28, 1995.

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, CT 06360.

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

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

Date of application for amendment: July 17, 1995.

Brief description of amendment: The amendment revises the Technical

Specifications pertaining to the plant air filtration and ventilation

systems to extend the surveillance frequencies that are now required to

be performed at least once per 18 months to specify that the

surveillances are to be performed at least once each refueling

interval.

Date of issuance: December 28, 1995.

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

within 90 days.

Amendment No.: 123.

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

Technical Specifications.

Date of initial notice in Federal Register: November 27, 1995 (60

FR 58402).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 28, 1995.

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, CT 06360.

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

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

Date of application for amendment: July 14, 1995.

Brief description of amendment: The amendment revises the frequency

of those surveillance requirements for the emergency core cooling

systems that now require that the surveillances be performed ``at least

once per 18 months'' to specify that the surveillances be performed

``at least once each refueling interval.''

Date of issuance: December 28, 1995.

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

within 90 days.

Amendment No.: 124.

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

Technical Specifications.

Date of initial notice in Federal Register: November 27, 1995 (60

FR 58402).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 28, 1995.

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, CT 06360.

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

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

California.

Date of application for amendments: September 29, 1995.

Brief description of amendments: The amendments added a one-time

footnote to the Technical Specifications related to the diesel

generator fuel oil storage and transfer system to permit each of the

existing storage tanks to be removed from service for up to 60 days so

they can be replaced with double walled tanks and piping that comply

with new California regulations.

Date of issuance: January 3, 1996.

Effective date: January 3, 1996, to be implemented within 30 days

of issuance.

Amendment Nos.: Unit 1--Amendment No. 109; Unit 2--Amendment No.

108.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 27, 1995 (60

FR 58403).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 3, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power

Plant, Unit 3, Humboldt County, California.

Date of application for amendment: October 8, 1993, as supplemented

October 28, 1994.

Brief description of amendment: This amendment revised the

Technical Specification by deleting Figure II-2, ``Restricted Area Per

10 CFR 20.3(a)(14)'' and by deleting the restricted area boundary line

from Figure V-3, ``HBPP Groundwater Monitoring System Wells.''

Date of issuance: December 21, 1995.

Effective date: This license amendment is effective as of the date

of its issuance and must be fully implemented no later than 30 days

from the date of issuance.

Amendment No.: 30.

Facility License No. DPR-7: This amendment revised the TS.

Date of initial notice in Federal Register: January 5, 1994 (59 FR

624).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 21, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Humboldt County Library, 1313

3rd Street, Eureka, California 95501.

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

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

Pennsylvania.

Date of application for amendments: March 31, 1995.

Brief description of amendments: The amendments incorporate a

change in the Station Technical Specifications for both units that

modifies the requirement in TS 4.4.4.3.a to have the pH of the reactor

coolant measured every 72 hours. The amendments add the clarification

that the pH measurement will be performed only when the coolant

conductivity is greater than 1.0 micro-mho/cm at 25 deg.C ( deg.77).

Date of issuance: January 3, 1996.

[[Page 1646]]

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

implemented within 30 days.

Amendment Nos.: 156 and 127.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 26, 1995 (60 FR

20522).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 3, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

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

Pennsylvania 18701.

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

Susquehanna Steam Electric Station, Unit 2, Luzerne County,

Pennsylvania.

Date of application for amendment: August 11, 1995.

Brief description of amendment: The amendment revises the Unit 2

Technical Specifications (TSs) to reestablish the original operability

requirements for the Neutron Flux function, and to delete the footnote

that was added to TS page 3/4 3-71 under Amendment No. 115, regarding

the length of time that the revised operability values were valid.

Date of issuance: January 3, 1996.

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

days.

Amendment No.: 128.

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

Technical Specifications.

Date of initial notice in Federal Register: September 13, 1995 (60

FR 47623).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 3, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

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

Pennsylvania 18701.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York.

Date of application for amendment: May 12, 1995.

Brief description of amendment: The amendment modifies the

Technical Specifications (TSs) to extend the surveillance test

intervals for the emergency service water system to support 24-month

operating cycles. Surveillance test interval extensions are denoted as

being performed ``every 24 months'' or ``at least once per 24 months''

consistent with the guidance provided in Generic Letter (GL) 91-04,

``Changes in Technical Specification Surveillance Intervals to

Accommodate 24-Month Fuel Cycle,'' dated April 2, 1991. The NRC staff

has determined that the proposed TS changes are in accordance with GL

91-04, and are therefore acceptable.

Date of issuance: December 21, 1995.

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

30 days.

Amendment No.: 230.

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 13, 1995 (60

FR 47623)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 21, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

South Carolina Electric & Gas Company, South Carolina Public Service

Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit

No. 1, Fairfield County, South Carolina.

Date of application for amendment: February 21, 1995, as

supplemented on August 31, 1995, and December 4, 1995.

Brief description of amendment: The amendment revises the Technical

Specifications (TS) support of the licensee's plan to implement the

revised 10 CFR Part 20, ``Standards for Protection Against Radiation.''

Also, several editorial changes to improve the clarity of the TS were

made.

Date of issuance: December 28, 1995.

Effective date: 90 days after issuance.

Amendment No.: 130.

Facility Operating License No. NPF-12. Amendment revises the

operating license.

Date of initial notice in Federal Register: March 29, 1995 (60 FR

16200). Renoticed on September 27, 1995 (60 FR 49946) due to changes in

the licensee's proposed no significant hazards consideration analysis

that were included in the August 31, 1995 supplemental letter. The

December 4, 1995 letter provided supplemental information that did not

change the second proposed no significant hazards consideration. The

Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated December 28, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri.

Date of application for amendment: June 21, 1994, as supplemented

by letter dated October 23, 1995.

Brief description of amendment: The amendment revises Technical

Specification (TS) 6.5.1, 6.5.2 and 6.5.3 to relocate the review and

audit requirements of the On-site Review Committee (ORC) and the

Nuclear Safety Review Board (NSRB) to the Operational Quality Assurance

Manual (OQAM). In addition, the amendment deletes reference to the

Manager, Nuclear Safety and Emergency Preparedness, in TS 6.2.3. The

Index is revised to reflect the relocations.

Date of issuance: December 26, 1995.

Effective date: December 26, 1995, to be implemented within 30 days

from the date of issuance.

Amendment No.: 107.

Facility Operating License No. NPF-30. The amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45036) and November 27, 1995 (60 FR 58406). The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

December 26, 1995.

No significant hazards considera

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