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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from January 7, 1994, through January 21, 1994.

The last biweekly notice was published on January 19, 1994 (59 FR

2859).

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 P-223, Phillips Building, 7920 Norfolk Avenue,

Bethesda, 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 20555.

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

intervene is discussed below.

By March 4, 1994, 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 20555 and at the local

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

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

date, the Commission or an Atomic Safety and Licensing Board,

designated by the Commission or by the Chairman of the Atomic Safety

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

and the Secretary or the designated Atomic Safety and Licensing Board

will issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, 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 20555, 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 20555, and at the local public document

room for the particular facility involved.

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

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

Date of amendment request: November 4, 1993

Description of amendment request: The proposed amendment revises

Technical Specification 6.13.1 to provide use of alarming dosimeters in

high radiation areas. This change includes newly revised 10 CFR Part 20

requirement references and is consistent with NUREG-1413, Standard

Technical Specifications - Westinghouse Plants, Specification 5.11.1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. This change could involve a reduction in

personnel radiation exposure by utilizing alarming dosimeters. This

change does not involve any plant systems or components which could

increase the probability of an accident. Therefore, there would be

no increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. This change could reduce the possibility of an accidental

overexposure by alerting personnel when their maximum allowable

exposure has been received. This change does not involve any plant

systems or components. 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 amendment does not involve a significant

reduction in the margin of safety. Due to the nature of this

proposed change, it is not related to any plant system. Therefore,

the proposed changes do not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

Home and Fifth Avenues, Hartsville, South Carolina 29550

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: S. Singh Bajwa

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

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

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

Will County, Illinois

Date of amendment request: August 27, 1993

Description of amendment request: The proposed amendments would

revise the Braidwood Station, Units 1 and 2, and Byron Station, Units 1

and 2, Technical Specifications (TS) regarding inspection requirements

for pipe snubbers. The proposed changes implement Generic Letter (GL)

90-09, and would affect the surveillance requirements of TS section

4.7.8 and the bases for these requirements, section 3/4.7.8.

Specifically, the amendment would change the existing inspection

periods, visual inspection acceptance criteria, and functional test

requirements. Additionally, there would be changes to the bases to

include reference to GL 90-09, and other editorial changes would be

made.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The amended surveillance requirement adds a table that addresses

the maximum number of snubber failures that can be tolerated prior

to reducing the inspection interval. This number is a function of

the population size of a particular type of snubber. The revised

requirement will allow the inspection intervals to be compatible

with the 24 month fuel cycles, and provisions are included to extend

the inspection interval up to 48 months. A provision is included to

allow an evaluation to determine operability to justify continued

operation with a snubber that is unacceptable.

The purpose of the amendment request is to provide for

alternative inspection intervals that take the size of the

population of a snubber type into account. The proposed change

provides the same confidence level and allows snubber inspection and

corrective action to be performed during refueling outages. This

allows the plant to avoid a mid-cycle outage due to a small number

of snubber failures.

The proposed change allows for a small percentage of snubbers in

each category to fail the required visual examination without

adjusting the inspection frequency. If a statistically significant

percentage of snubbers fail, the visual examination inspection

interval is reduced based on the percentage of failed snubbers.

The proposed change has no direct or indirect impact on

reactivity management activities.

The change is not expected to have an impact on equipment

failures. Any snubbers that fail to meet the visual examination

acceptance criteria are either functionally tested in the as-found

condition to verify continued acceptability, or an evaluation is

performed to demonstrate the acceptability of continued operation

with an unacceptable snubber. No new equipment is being introduced

and no systems are operated in a configuration that has not been

evaluated, so no new failure modes are introduced.

The affected transients are the design basis earthquake and the

spectrum of event initiating transients, with the capability of

imposing significant dynamic loads or otherwise which impact the

structural integrity of the Reactor Coolant System (RCS).

The snubbers are installed to ensure the structural integrity of

the RCS and required support systems. Their failure is passive in

nature. The probability of a transient initiating event occurring is

unrelated to the existence or condition of equipment that is

designed to perform a mitigating function. The snubbers are

installed to ensure an acceptable system response to a dynamic load,

and their availability does not impact the frequency of occurrence

of earthquakes or other transients resulting in significant dynamic

loading.

The revised testing provisions are designed to allow some

flexibility while still maintaining a high probability that the

installed snubbers will be capable of performing their intended

function when required. The revised surveillances appropriately

consider the size of the population of a particular type of snubber,

and are sufficient to ensure the consequences of an accident will be

unchanged when the revised requirements are implemented. By

maintaining a statistically high level of confidence in the function

of the plant's snubbers, the system response to transient initiating

events will be as designed and thus, the off-site dose projected to

occur of any affected transient will remain acceptably low.

As previously stated, the revised surveillance provides a high

confidence that the affected systems will remain intact and

functional. Evaluation of the effects of operating with a degraded

snubber is required to ensure that adequate margin exists to support

continued plant operation. If this evaluation cannot adequately

justify continued operation, the appropriate action statement will

be applied. These provisions are sufficient to assure that the

probability of an equipment malfunction will not increase.

The consequences of equipment malfunction will not increase.

Sufficient redundancy exists to accommodate the complete failure of

one train of required equipment. The requisite electrical and

physical separation are sufficient to ensure that the redundant

train remains unaffected. This redundancy is adequate to ensure that

the undetected failure of a snubber will not have a severe impact on

overall system response to a transient.

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

or different kind of accident from any accident previously

evaluated.

The possibility of a new or different type of accident is not

created by this change. No new or different equipment is being

introduced, and no system will be operated in a different

configuration without first having the effects of the new

configuration evaluated. The new configuration would be system and/

or plant operation with a snubber installed that has failed its

visual examination. The required evaluation must be sufficient to

provide confidence that continued operation is acceptable;

otherwise, the provisions of the action statement will be observed.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

No reduction in the margin of safety will occur as a result of

this proposed change. As previously described, the controls in place

will provide a high confidence the affected systems will continue to

be functional. No significant increase in the rate of occurrence of

undetected inoperable snubbers is expected to occur, and the

allowable failures prior to applying an increased test frequency is

still a small percentage of the total snubber population.

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: For Byron, the Byron Public

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

Braidwood, the Wilmington Township Public Library, 201 S. Kankakee

Street, Wilmington, Illinois 60481

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: James E. Dyer

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

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

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

Will County, Illinois

Date of amendment request: September 2, 1993, supplemented by

letter dated January 7, 1994.

Description of amendment request: The proposed amendment would

revise the Braidwood Station, Units 1 and 2, and Byron Station, Units 1

and 2, Technical Specifications (TS) to allow replacement of the 125

Volt DC Gould batteries with the new 125 Volt DC AT&T batteries and

rephrase their design duty cycle. In addition, the proposed amendment

would revise the batteries crosstie loading limitations and the

crosstie breaker limitations. The associated Bases would also be

revised to discuss the purpose for the crosstie limitations and to

discuss design duty cycle 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 change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The replacement AT&T battery has been selected to meet or exceed

the design, functional, and operational requirements of those of the

present Gould battery, including crosstie load limitations. The

crosstie breaker limitation change to allow crosstie between two

shutdown units is consistent with the Safety Evaluations issued with

Technical Specification Amendment 5 for both Braidwood and Byron

Stations. The remaining changes are administrative in nature or

provide clarification to maintain consistency with other Technical

Specifications and the Standard Technical Specifications.

The overall design, function, and operation of the DC system and

equipment has not been altered by these changes. The proposed

changes do not affect any accident initiators or precursors and do

not alter the design assumptions for the systems or components used

to mitigate the consequences of an accident as analyzed in UFSAR

Chapter 15. Therefore, there is no increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The replacement AT&T battery will provide the same functions as

those of the present Gould battery and will be operated with the

same types of operational controls. These limits will include

battery float terminal voltage, individual cell voltage and

electrolyte specific gravity, and crosstie loading. Crosstie

conditions are allowed under the present Technical Specifications.

The remaining changes are administrative in nature or provide

clarification to maintain consistency with other Technical

Specifications and the Standard Technical Specifications.

The DC system and its equipment will continue to perform the

same functions and be operated in the same fashion. The proposed

change does not create any new or common failure modes. The proposed

changes do not introduce any new accident initiators or precursors,

or any new design assumptions for the systems or components used to

mitigate the consequences of an accident. Therefore, the possibility

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

evaluated has not been created.

C. The proposed change does not involve a significant reduction

in a margin of safety.

The replacement AT&T battery will meet or exceed the design,

functional, and qualification requirements of those of the present

Gould batteries. The proposed Technical Specification limitations

for the AT&T battery are derived from the same methodology and

margins as those for the Gould battery. Increasing the crosstie

loading limit takes advantage of the larger AT&T battery capacity

with its increased design margin. The proposed change to the

crosstie loading limit will continue to conservatively envelop the

postulated design requirements. The remaining changes are

administrative in nature or provide clarification to maintain

consistency with other Technical Specifications and the Standard

Technical Specifications.

The inherent design conservatism of the DC system and its

equipment has not been altered. The DC system and its equipment will

continue to be operated with the same degree of conservatism.

Therefore, there is no 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: For Byron, the Byron Public

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

Braidwood, the Wilmington Township Public Library, 201 S. Kankakee

Street, Wilmington, Illinois 60481

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: James E. Dyer

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: October 28, 1993

Description of amendment request: The proposed amendment would

revise the Emergency Core Cooling System (ECCS) injection valve stroke

times and ECCS response times for Motor-Operated Valve (MOV)

modifications that increase injection valve stroke times.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because:

The probability of an accident previously evaluated will not

increase as a result of this change, because the only modification

being performed is to the stroke times for the LPCS [Low Pressure

Core Spray System], LPCI [Low Pressure Coolant Injection System],

and HPCS [High Pressure Core Spray System] injection valves.

Changing the opening or closing time of the injection valves for

these ECCS systems does not cause any accident previously evaluated

to occur. Therefore, modifying their stroke times will not increase

the probability of occurrence for any accident previously evaluated.

The consequences of a LOCA [Loss of Coolant Accident] are not

significantly increased and do not exceed the previously accepted

licensing criteria for this accident. GE [General Electric Company]

has calculated the revised licensing basis PCT [Peak Centerline

Temperature] for LaSalle Station to be 1260 deg.F, which is well

below the 2200 deg.F criterion of 10 CFR 50.46 and Section 15.6.5 of

NUREG-0800 (Standard Review Plan). The acceptance criteria for

cladding oxidation, metal-water reaction (hydrogen generation),

coolable geometry and long-term cooling also continue to be met with

the increased valve stroke times.

GE has performed sensitivity analysis justifying the continued

applicability of previous analyses for Anticipated Transients

Without Scram (ATWS), containment analyses, off-site dose (Main

Steamline Break Outside Containment), and HPCS-related transients

(Loss of Feedwater Flow). Other events are not affected because

these systems are not assumed to function.

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

or different kind of accident from any accident previously evaluated

because:

The only modification is to increase the stroke time of the

injection valves for LPCS, LPCI, and HPCS. This does not result in

any changed component interactions, other than to increase the

affected ECCS response times. The injection valves will still

provide the function for which they were designed. Since the systems

will continue to function as intended, the proposed 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 a margin of safety because:

While the calculated licensing basis PCT is larger than that

previously calculated with the current valve stroke times, the new

PCT remains far below the 2200 deg.F licensing acceptance limit

associated with a LOCA. This limit has been previously evaluated as

providing a sufficient margin of safety. All other LOCA licensing

limits also continue to be met with the increased stroke times. For

other accidents and transients, the increased stroke times have a

negligible effect on the results, so the margin of safety is

preserved.

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: Public Library of Illinois

Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: James E. Dyer

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan

Date of amendment request: April 15, 1992, as modified by letters

dated December 8, 1992, and June 25, 1993.

Description of amendment request: The proposed amendment would

revise the provisions in the Technical Specifications to incorporate

Generic Letter 90-06, ``Resolution of Generic Issue 70, `Power-Operated

Relief Valve and Block Valve Reliability,' and Generic Issue 94,

`Additional Low-Temperature Overpressure Protection for Light-Water

Reactors,''' power-operated relief valve (PORV) requirements for power

operation, and to modify the primary coolant system (PCS) overpressure

protection specification venting 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:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed Technical Specification requiring Operability of

the PORVs and their block valves does not alter plant operation or

configuration in any way. It is current practice to maintain these

valves in an Operable condition to meet the requirements of existing

Specification 3.1.8, which is applicable when below 430 deg.F. The

effect of the proposed changes is to extend the applicability of the

Operability requirement for these valves. The addition of PORV

Operability requirements when at Hot Standby and above would not

involve a significant increase in the probability or consequence of

an accident previously evaluated.

Replacing the requirements to vent the PCS through a 1.3 square

inch vent with a vent capable of relieving 167 gpm at a pressure

less than the Appendix G limits will not significantly increase the

probability or consequences of an overpressurization event

occurring. The 1.3 square inch vent area in Technical Specification

3.1.8 was intended to be a means of protecting the Primary Coolant

System (PCS) from exceeding the limit of the 10 CFR [Part] 50,

Appendix G, curve following an overpressure transient. Analysis has

shown that manual vent valves PC-514 and PC-515 will provide a

relief capacity of 167 gpm at a PCS pressure of approximately 115

psig, well below the minimum 331 psig limit (Appendix G curve limit

for a 40 deg.F/hr heat-up). This relief capacity will protect the

PCS against a pressure transient caused by a maximum charging/

letdown imbalance coincident with a 40 deg.F/hr PCS heat-up rate and

a 60 deg.F/hr pressurizer heat-up rate.

Two other pressure transients, a High Pressure Safety Injection

(HPSI) pump start and a Primary Coolant Pump (PCP) start, are also

precluded. With the PCS in a vented and depressurized state, the PCS

would be below 212 deg.F Existing technical specifications require

both HPSI pumps to be rendered inoperable below 260 deg.F and, with

the system depressurized, normal operating procedures prohibit a PCP

start due to insufficient pump net positive suction head (NPSH).

Therefore, the 1.3 square inch requirement can be replaced with

a requirement to have a vent capable of relieving 167 gpm at a PCS

pressure less than or equal to the Appendix G limit with no

significant increase in the probability or consequences of an

overpressurization event occurring.

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

from any previously evaluated.

The addition of PORV Operability requirements when at Hot

Standby and above will not alter plant operation or configuration.

It will not alter any equipment or analyses. Therefore the addition

of these PORV Operability requirements will not create the

possibility of a new or different kind of accident from any

previously evaluated.

The new technical specifications requirements for PCS vent

capacity will provide an equivalent or better, overpressure

protection as compared to the existing requirement. No analysis has

been found that shows that the existing 1.3 square inch vent area

will protect the PCS from exceeding the Appendix G curve limit.

However, analysis has been developed which shows that manual vent

valves PC-514 and PC-515 will provide adequate relief capacity,

maintaining PCS pressure within the 10 CFR [Part] 50, Appendix G,

limits. Furthermore, two other pressure transients, a HPSI pump

start and a PCP start, are also precluded by either existing

technical specifications or normal plant operating procedures.

Another related analysis has shown that relief valve RV-3164,

the Low Pressure Safety Injection (LPSI) pumps, the LPSI pump seals,

and the system piping of the shutdown cooling system have the

capability of providing adequate overpressure protection to the

shutdown cooling system.

The addition of the manual vent valves do not introduce a vent

path where a vent path had not previously existed. Therefore, the

possibility of an accident of a new or different type, than

previously evaluated in the FSAR, will not be created.

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

The margin of safety will not be reduced by the proposed

Technical Specifications changes. The extension of PORV Operability

requirements has no effect on any margin of safety. The previous

requirement assumed that a vent with an equivalent flow area as the

original PORV would provide the same relief as the PORV itself and

gave no consideration to how that flow area should factor in system

losses or vent location. The new technical specification requirement

offers a means to ensure the PCS will be protected against all

achievable overpressure transients for the system configuration,

with analyses to support it.

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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power

Company, 212 West Michigan Avenue, Jackson, Michigan 49201

NRC Project Director: A. Randolph Blough, Acting

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: December 16, 1993

Description of amendment request: The proposed amendment clarifies

the requirements for maintaining secondary containment integrity when

one or more Reactor Building Ventilation supply and exhaust valves are

declared inoperable. The technical specification (TS) change adds a new

Limiting Condition for Operation, Basis Statement, and Surveillance

Requirements for these isolation valves. The change revises TS

definition 1.14c, adds new Specifications 3.5.B.2, 3.5.B.3, 4.5.R, and

a Basis statement to TS 3.5, edits T.S. 3.5.B.1.1. It also renumbers TS

3.5.B.2 through 3.5.B.4, to 3.5.B.4 through 3.5.B.7. It also revises

specification references within to reflect new specification numbers.

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:

GPU Nuclear has determined that operation of the Oyster Creek

Nuclear Generating Station in accordance with the proposed Technical

Specifications does not involve a significant hazard. The changes do

not:

1. Involve a significant increase in the probability or the

consequence of an accident previously evaluated.

The failure of any component in the [Reactor Building

Ventilation System] RBVS was not considered as a credible initiating

event for a design basis accident. However, the RBVS is designed to

mitigate the consequences of a potential radiological release by the

isolation of all supply and exhaust ducts to the environs. Since the

failure of the RBVS was never considered as one of the initiators of

an accident, this proposed change cannot increase the probability of

occurrence of an accident. During the proposed Limiting Condition

for Operation (LCO), the supply or exhaust duct will be isolated

within 8 hours by one isolation valve secured in its post accident

design position. Since the duct can perform its post accident design

function (isolation), there is no increase in the consequences of an

accident.

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

from any previously evaluated.

The design function of the RBVS automatic isolation valves is to

isolate the ducts which penetrate the Reactor Building or Secondary

Containment during a radiological release. During the LCO, the duct

will be isolated within 8 hours by one isolation valve secured in

the closed position. Since the duct will be isolated, this change

will not create a possibility for an accident or malfunction of a

different type than previously identified.

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

If a RBVS automatic isolation valve (supply) is declared

inoperable, the proposed LCO would allow continued plant operation

with that supply duct isolated. Since the RBVS can still perform its

design function (redundant ductwork) under normal plant and design

accident conditions, there is no reduction in the margin of safety.

For an inoperable isolation valve in the exhaust duct, the exhaust

duct will be isolated within 8 hours by one isolation valve secured

in the closed position. Further, the RBVS and the [Standby Gas

Treatment System] SGTS will be aligned for an accident condition

with no 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: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, New Jersey

08753

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

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

NRC Project Director: John F. Stolz

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: November 17, 1993

Description of amendment requests: The proposed amendments would

modify the Technical Specifications to allow a portion of the Waste Gas

Holdup System (WGHS) Explosive Monitoring System to be inoperable for

160 days on a one-time basis. This is to allow replacement of the Waste

Gas Oxygen Analyzer.

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:

We have evaluated the proposed T/S changes and have determined

that the changes should involve no significant hazards

consideration. Operation of the Cook Nuclear Plant in accordance

with the proposed amendment will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The purpose of the hydrogen and oxygen monitors is to

measure the concentrations of these gases in the WGHS to ensure that

the gas mixture is non-flammable. We can accomplish this purpose and

ensure safe operation of the WGHS by operating the system in the

proposed manner. During the equipment replacement interval, we will

be assuming that the hydrogen concentration is above the flammable

limit (4%). The limiting factor then is the oxygen concentration at

which hydrogen and oxygen become flammable. The existing hydrogen

monitor will be in continuous operation to verify the hydrogen

level. The information provided by the oxygen monitor being replaced

is not essential to the safe operation of the WGHS since it is

redundant to information provided by the remaining oxygen monitor.

The only difference between the two (2) oxygen monitors is that the

one being replaced provides an automatic isolation of the waste gas

decay tank when the oxygen concentration reaches 3%. The isolation

of the waste gas decay tanks will be performed manually during the

replacement. In the event the remaining oxygen monitor becomes

inoperable, we will follow the currently approved T/Ss. Since

operation of the WGHS in the manner we have proposed will ensure

that the purpose of the oxygen and hydrogen monitors is fulfilled

and safe operation of the WGHS is maintained, the proposed change

will not involve a significant increase in the probability or

consequences of a previously analyzed incident.

The proposed change to the Automatic Gas Analyzer (QC-31) tag

number to QC-1400 will not reduce in any way requirements or

commitments in the existing T/Ss. The proposed change will eliminate

confusion of spare parts of the new analyzer panel installed in

1990.

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

accident from any previously analyzed.

The proposed amendment does not create the possibility of a new

or different kind of accident from any previously evaluated. During

the replacement of the monitors the waste gas holdup system will

continue to operate normally. The proposed method of operation will

ensure that the oxygen and hydrogen gas mixture is non-flammable.

For this reason, operating the explosive gas monitoring system in

the proposed manner will not place the plant in a new or unanalyzed

condition. Therefore, we believe that this change will not introduce

a new or different kind of accident than previously analyzed.

The proposed editorial change will not create the possibility of

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

because these changes will not place the plant in a new or

unanalyzed condition.

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

The proposed amendment does not involve a significant reduction

in the margin of safety. The remaining oxygen monitor will be

available to maintain the oxygen concentration below the limit

required for hydrogen flammability in oxygen. In addition, the

oxygen grab samples will provide redundant information and will

serve as a check of the monitor's readings. If the remaining monitor

becomes inoperable, we will follow the actions of our current T/Ss.

During the equipment replacement period, we will be assuming that

the hydrogen concentration is above the flammable limit (4%). This

will then make the oxygen level the controlling parameter in a

possible flammable combination of oxygen and hydrogen. The existing

hydrogen monitor will be in continuous operation to verify the

hydrogen level. These proposed interim measures will not

significantly affect our ability to maintain the hydrogen and oxygen

concentration within the limits to prevent flammability. Therefore,

we believe that operation of the system in this manner does not

involve a significant reduction in a margin of safety.

The proposed editorial change will not involve a significant

reduction in margin of safety, because all accident analyses and

nuclear design bases remain unchanged.

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: A. Randolph Blough, Acting

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 20, 1993

Description of amendment requests: The proposed amendments would

revise the Technical Specifications to change Train A and B Emergency

Loads from 8 hour to composite 4 hour, delete a load on the Train B

batteries load list, and revise the operational loads on the Train N

batteries.

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:

Per 10 CFR 50.92, a proposed amendment does not involve a

significant hazards consideration if the change does not:

(1) involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

accident from any previously analyzed.

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

Criterion 1

The change is proposed to comply with the LOCA/LOOP [loss-of-

coolant accident/loss-of-offsite power] and SBO [Station Blackout]

requirement for Cook Nuclear Plant for battery testing. The

composite test as addressed above meets these requirements for four

hour test profiles. This change is consistent with the UFSAR three

hour LOCA/LOOP and NUMARC 87-00 Station Blackout Rule four hour.

Based on these considerations, the proposed change does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

Criterion 2

The change only addresses the battery profile test which meets

both LOCA/LOOP and SBO for Cook Nuclear Plant. No specific physical

or operational changes to the plant will occur due to this change.

Thus, the change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Criterion 3

The proposed change revises the battery profile test table from

an eight hour to a four composite test which complies with both

LOCA/LOOP and SBO as defined for Cook Nuclear Plant. Therefore, the

proposed change does not involve a significant reduction in a margin

of safety.

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

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

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

amendment 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: A. Randolph Blough, Acting

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 22, 1993

Description of amendment requests: The proposed amendments would

revise the Technical Specifications section addressing steam generator

stop valves making it more consistent with the revised Standard

Technical Specifications and clarifying certain surveillance

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:

Per 10 CFR 50.92, a proposed change does not involve a

significant hazards consideration if the change does not:

1. involve a significant increase in the probability or

consequences of an accident previously evaluated,

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

from any accident previously evaluated, or

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

Criterion 1

The limiting conditions for operation involving the steam

generator stop valves are not altered by this proposed change. The

surveillance requirements are lessened for Unit 2 in that valve

stroke timing does not have to be performed on valves that are

closed. This is consistent with the wording of the Unit 1 T/S, and

reflects the fact that, when closed, the valves are already in the

position required by the assumptions in the safety analysis and

therefore stroke timing is not necessary. The remaining changes are

consistent with NUREG 1431, and as such, have already been found

acceptable by the NRC. Therefore, it is concluded that the proposed

amendment does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

Criterion 2

No changes to the limiting conditions for operation of the steam

generator stop valves are proposed as part of this amendment

request. The proposed changes do not involve any physical changes to

the plant. The changes will allow operation in Modes 2 and 3 with

more than one steam generator stop valve inoperable. However,

inoperable valves must be closed and their closure periodically

reverified. When closed, the valves are already in the position

required by the assumptions in the safety analysis. Thus, the

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

kind of accident from any accident previously evaluated.

Criterion 3

The limiting conditions for operation involving the steam

generator stop valves are not altered by this proposed change. The

surveillance requirements are lessened for Unit 2 in that valve

stroke timing does not have to be performed on valves that are

closed. This is consistent with the wording of the Unit 1 T/S, and

reflects the fact that, when closed, the valves are already in the

position required by the assumptions in the safety analysis and

therefore stroke timing is not necessary. The remaining changes are

consistent with NUREG 1431, and, as such, have already been found

acceptable by the NRC. Therefore, 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 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: A. Randolph Blough, Acting

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile

Point Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: December 22, 1993

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.4.4.e (Emergency Ventilation

System). TS 3.4.4.e currently permits fuel handling operations to

continue during refueling for up to 7 days with one circuit of the

emergency ventilation system inoperable provided all active components

of the other emergency ventilation system circuit are operable. The

proposed revision would permit fuel handling operations to continue

during refueling beyond 7 days with one circuit of the emergency

ventilation system inoperable provided the remaining emergency

ventilation system circuit is operable and in operation. The licensee

stated that the proposed revision is consistent with recently issued

Amendment No. 47 to the Nine Mile Point Unit 2 TSs and with the NRC's

Improved Standard Technical Specifications, NUREG-1433.

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 operation of Nine Mile Point Unit 1 [NMP1], in accordance

with the proposed amendment, will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The Emergency Ventilation System responds to a release of

radioactivity to the secondary containment by maintaining a negative

pressure in secondary containment and by providing a filtered

elevated release. The proposed change to LCO [Limiting Condition for

Operation] 3.4.4.e would allow continuation of refueling beyond

seven days with one emergency ventilation circuit inoperable

provided the operable emergency ventilation circuit is in operation.

A plant specific PRA [Probabilistic Risk Assessment] was performed

to evaluate the probability of a fuel bundle drop event resulting in

a need to start the Emergency Ventilation System with a concurrent

failure of the Emergency Ventilation System that would result in an

unfiltered ground level release under the current Technical

Specifications and the proposed change. The results of this

assessment indicate that the probability is not significantly

increased. In addition, the order of magnitude of the probability of

such a release, under the current or proposed Technical

Specifications, is very small, i.e., 10-6. This amendment

requires no physical changes to NMP1. Therefore, the proposed

changes to the Technical Specifications do not significantly

increase the probability of an accident previously evaluated.

Section XV.C.3 of the UFSAR [Updated Final Safety Analysis

Report] evaluates a fuel bundle drop accident. The radiological

consequences of this accident are within the guidelines of 10 CFR

Part 100. The UFSAR radiological evaluation takes credit for the

operation of an emergency ventilation circuit in mitigating the

consequences of this accident. During refueling with one emergency

ventilation circuit inoperable for more than seven days, the

proposed Technical Specification change would require that an

operable emergency ventilation circuit be placed in operation. With

an operable emergency ventilation circuit operating prior to a fuel

bundle drop accident, the radiological consequences of this accident

remains bounded by the current UFSAR evaluation. Therefore, from a

radiological perspective, the proposed Technical Specification

change is bounded by the current radiological evaluation in the

UFSAR. Therefore, the Technical Specification change does not

significantly increase the consequences of a previously evaluated

accident.

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

This amendment does not involve any accident precursors or

initiators. During an accident involving the release of

radioactivity to the secondary containment atmosphere an operable

emergency ventilation circuit would already be running and

performing its safety function. The operating status of a running

emergency ventilation circuit, which was manually started, would be

unaffected by the receipt of an automatic start signal due to the

detection of high radiation in secondary containment. Accordingly,

the proposed Technical Specification change will not create the

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

previously evaluated.

The operation of Nine Mile Point Unit 1, in accordance with

proposed amendment, will not involve a significant reduction in a

margin of safety.

The current Technical Specifications, LCO 3.4.4, provide a

margin of safety by requiring both emergency ventilation circuits to

be operable during a refueling condition. With one emergency

ventilation circuit inoperable, the current Technical Specifications

allow continuation of refueling for up to seven days, at which time

refueling must be stopped. This Technical Specification requirement

ensures that an emergency ventilation circuit will be available to

provide a filtered release to the environment during an accident

which could result in the release of radioactivity to the secondary

containment atmosphere.

The proposed change to LCO 3.4.4.e would allow continuation of

refueling beyond seven days with one emergency ventilation circuit

inoperable provided the operable emergency ventilation circuit is in

operation. By placing the remaining operable emergency ventilation

circuit in operation, active single failures associated with its

startup have been eliminated. These eliminated failures include

automatic initiation instrumentation, relaying logic, breaker

operation, fan startup and valve operation. With an operable

emergency ventilation circuit in operation, its safety function is

being performed. In addition, the status of the operating emergency

ventilation circuit is indicated in the control room. Therefore, the

running, operable emergency ventilation circuit provides a level of

safety comparable to two non-running, operable emergency ventilation

circuits.

Based upon the above analysis, the margin of safety is not

significantly reduced by allowing refueling to continue beyond seven

days with one emergency ventilation circuit inoperable since the

operable emergency ventilation circuit is in operation.

These changes are consistent with Amendment No. 47 for Nine Mile

Point Unit 2 and with the Improved Standard Technical

Specifications, NUREG-1433.

Accordingly, as determined by the analysis above, this proposed

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: Reference and Documents

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

York 13126.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &

Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Project Director: Robert A. Capra

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile

Point Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: December 27, 1993

Description of amendment request: The proposed amendment would

relocate Technical Specification (TS) Tables 3.2.7, ``Reactor Coolant

Isolation Valves,'' and 3.3.4, ``Primary Containment Isolation

Valves,'' from TS 3.2.7/4.2.7 and 3.3.4/4.3.4, respectively, to a plant

procedure overning lists removed from TSs per Generic Letter (GL) 91-

08, ``Removal of Component Lists from Technical Specifications.'' The

plant procedure would be subject to the requirements specified in the

Administrative Controls section of the Nine Mile Point Nuclear Station

Unit No. 1 (NMP-1) TS. The proposed amendment would also make

conforming changes to the TS Bases. These lists of valves will continue

to be included in the NMP-1 Updated Final Safety Analysis Report

(FSAR). The licensee stated that the proposed changes would be

consistent with NRC staff guidance issued in GL 91-08.

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 operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

1.The proposed amendment does not involve a physical change to

any system, structure, or component that affects the probability or

consequences of any accident or malfunction of equipment important

to safety.

Relocation of the component lists to plant procedures and the

Updated FSAR is in accordance with Generic Letter 91-08. This change

does not alter the existing operability or surveillance requirements

for the components to which they apply. The proposed changes are

under the change control provisions in Section 6.0, ``Administrative

Controls,'' of the Technical Specifications. The changes associated

with the Bases for Specification[s] 3.2.7/4.2.7 and 3.3.4/4.3.4 are

consistent with the issuance of prior license amendments. Since the

proposed amendment does not affect the operation or testing of any

plant systems or components, it will have no impact on the

probability or consequences of accidents or malfunctions previously

evaluated.

2.The operation of Nine Mile Point Unit 1, in accordance with

the proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes to Technical Specification 3.2.7/4.2.7,

3.3.4/4.3.4 and Bases do not introduce any new modes of plant

operation or new accident precursors, involve any physical

alterations to plant configurations, or make changes to system

setpoints which could initiate a new or different kind of accident.

The proposed changes relocate Reactor Coolant Primary Containment

Isolation Valve Tables 3.2.7 and 3.3.4 into a procedure governing

controlled lists removed from TS per Generic Letter 91-08 under the

change control provisions in Section 6.0, ``Administrative

Controls,'' of the Technical Specifications. The testing associated

with these valves remains unchanged, therefore, it will not affect

system or component operability. In addition, the removal of generic

reference to the 60 second closure time is consistent with

previously issued license amendments and has no impact on either the

Limiting Condition for Operation or Surveillance Requirement.

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

new or different kind of accident from any previously evaluated.

3. The operation of Nine Mile Point Unit 1, in accordance with

the proposed amendment, will not involve a significant reduction in

a margin of safety.

The Technical Specification Limiting Conditions for Operation

and Surveillance Requirements for the valves listed in Tables 3.2.7

and 3.3.4 are not being altered. The valve lists will be

incorporated into a procedure governing controlled lists removed

from TS per Generic Letter 91-08. This is controlled by Section 6.0,

``Administrative Procedures.''

In addition, removal of generic reference to the 60 second

closure time is consistent with previously issued license amendments

and has no impact on either the Limiting Conditions for Operation or

Surveillance Requirements. Therefore, the proposed changes will not

involve a significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Mark J. Wetterhahn, Esquire, Winston &

Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Project Director: Robert A. Capra

North Atlantic Energy Service Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: October 4, 1993

Description of amendment request: The proposed amendment would

delete the emergency diesel-generator engine speed specification from

Surveillance Requirement (SR) 4.8.1.1.2a.5 and would replace the diesel

engine speed requirement with an electrical frequency requirement in SR

4.8.1.1.2g. Currently, SR 4.8.1.1.2a.5 specifies both a minimum engine

speed and a nominal electrical frequency and acceptable deviation from

the nominal value. SR 4.8.1.1.2g currently specifies only a minimum

engine speed. The specified minimum engine speed is not consistent with

the acceptable frequency deviation below the nominal value.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below.

A. The changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated (10

CFR 50.92(c)(1)) because the proposed changes merely delete the

specification of emergency diesel-generator minimum engine speed

while retaining or substituting the specification of electrical

frequency to be attained by the emergency diesel-generator. The

diesel engine and generator are directly coupled and must rotate at

the same speed, therefore, speed and frequency are directly related

and specification of acceptance values for both parameters is

redundant. Furthermore, electrical frequency, not engine speed, is

the parameter of interest with regard the ability of the system to

power emergency loads. The proposed changes do not affect the

adequacy of the surveillance test or the reliability of the system

to power emergency loads, and do not involve any physical changes to

facility structures, systems, or components. Therefore, since the

reliability of the emergency diesel-generators will not be reduced,

the probability or consequences of any accident previously evaluated

is not increased.

B. The changes do not create the possibility of a new or

different kind

of accident from any accident previously evaluated (10 CFR

50.92(c)(2)) because no physical changes to facility structures,

systems, or components are involved and they do not affect the

manner by which the facility is operated.

C. The changes do not involve a significant reduction in a

margin of safety (10 CFR 50.92(c)(3)) because the proposed changes

do not affect the manner by which the facility is operated or

involve changes to equipment or features which affect the

operational characteristics of the facility. 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: Exeter Public Library, 47

Front Street, Exeter, New Hampshire 03833.

Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One

International Place, Boston Massachusetts 02110-2624.

NRC Project Director: John F. Stolz

North Atlantic Energy Service Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: October 28, 1993

Description of amendment request: The proposed amendment would

implement 13 of the 47 line item Technical Specification (TS)

improvements recommended by Generic Letter 93-05. Most of the proposed

changes would revise the allowable time intervals for performing

certain Surveillance Requirements (SR) on various plant components

during power operation or would delete the requirement entirely or

under certain conditions. One proposed change would modify testing

requirements identified in an ACTION statement. The specific changes

are as follows:

1. SR 4.1.3.1.2 - The allowable interval between tests to

demonstrate the operability of any partially or fully withdrawn control

rod would be increased to 92 days from 31 days.

2. SR 4.6.4.1 - The allowable interval between tests to demonstrate

the operability of the hydrogen monitors by performing an Analog

Channel Operational Test (ACOT) would be increased to 92 days from 31

days, and by performing a Channel Calibration to every refueling outage

from 92 days on a staggered basis.

3. SR 4.3.2.1, Table 4.3-2, Functional Unit 3.c.4 and SR 4.3.3.1,

Table 4.3-3, Functional Units 1 through 6 - The allowable interval

between tests to demonstrate the operability of the radiation monitors

by performing an ACOT and Digital Channel Operational Test (DCOT) would

be increased to quarterly from monthly.

4. SR 4.4.6.2.2- The time the plant may be in Cold Shutdown before

Pressure Isolation Valve testing is required prior to entering Mode 2

would be increased to 7 days from 72 hours.

5. SR 4.4.11.1 - The allowable interval between tests to

demonstrate the operability of the Reactor Coolant System vent block

valves would be increased to cold shutdown from every 92 days.

6. SR 4.4.3.2 - The allowable interval between tests to verify

pressurizer heater capacity would be increased to each refueling outage

from 92 days.

7. SR 4.5.1.1.1 - The requirement to verify the boron

concentration of the accumulator contents after a volume increase of

1-percent or more would be removed under certain conditions.

8. SR 4.5.1.1.2 - The requirement to perform an ACOT and Channel

Calibration on accumulator water level and pressure instrumentation

would be deleted.

9. SR 4.5.2 - The requirement to visually inspect the containment

sump upon completion of each containment entry would be modified to

avoid unnecessary containment sump inspections when multiple

containment entries are made on the same day.

10. SR 4.6.2.1 - The allowable interval between tests to verify

that each containment spray nozzle is unobstructed would be increased

to every 10 years from 5 years.

11. SR 4.6.4.2 - The allowable interval between tests to

demonstrate operability of each hydrogen recombiner system would be

increased to each refueling interval from 6 months.

12. SR 4.7.1.2.1 - The allowable interval between tests of the

auxiliary and startup feedwater pumps would be increased to 92 days on

a staggered test basis from 31 days.

13. TS 3.8.1.1 - The ACTION statements would be changed so that

when the Limiting Conditions for Operation are not met due to:

a. Inoperability of one or two offsite power circuits the starting

of a diesel-generator would no longer be required;

b. Inoperability of one diesel-generator, starting of the remaining

diesel-generator would be required within 8 hours only under certain

conditions instead of within 24 hours under all conditions;

c. Inoperability of a diesel-generator and an offsite power

circuit, starting of the remaining diesel-generator would be required

within 8 hours only under certain conditions instead of within 24 hours

under all conditions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below.

A. The changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated (10 CFR

50.92(c)(1)) because the proposed changes either merely modify the

allowable intervals between certain surveillance tests, delete the SR,

or alter an ACTION statement with regard to required testing. The

safety functions of the related structures, systems, or components are

not changed in any manner nor are the reliabilities of any structure,

system, or component reduced by the revised surveillance or testing

requirements. The changes do not affect the manner by which the

facility is operated and do not change any facility design feature,

structure, system, or component. Since there is no change to the

facility or operating procedures, and the safety functions and

reliabilities of structures, systems, or components are not affected,

there is no affect upon the probability or consequences of any accident

previously analyzed.

B. The changes do not create the possibility of a new or different

kindof accident from any accident previously evaluated (10 CFR

50.92(c)(2)) because they do not change the facility or affect the

manner by which the facility is operated. The proposed changes merely

change certain surveillance or testing requirements.

C. The changes do not involve a significant reduction in a margin

of safety (10 CFR 50.92(c)(3)) because they do not affect the manner by

which the facility is operated or change equipment or features which

affect the operational characteristics of the facility.

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: Exeter Public Library, 47

Front Street, Exeter, New Hampshire 03833.

Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One

International Place, Boston Massachusetts 02110-2624.

NRC Project Director: John F. Stolz

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

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

Connecticut

Date of amendment request: December 17, 1993

Description of amendment request: The proposed amendment changes

the action statements for the limiting conditions associated with the

electrical power sources (Technical Specification 3.8.1.1).

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:

In accordance with 10 CFR 50.92, NNECO has reviewed the attached

proposed changes and has concluded that they do not involve a

significant hazards consideration. The basis for this conclusion is

that the three criteria of 10CFR50.92(c) are not compromised. The

proposed changes do not involve a significant hazards consideration

because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to rewrite the action statements for

Technical Specification 3.8.1.1 will decrease the wear on the EDGs

[emergency diesel generators] by reducing the number of required

starts. These changes will also allow adequate time for the

completion of manufacturer recommended EDG engine prelube and warmup

procedures. They ensure that the operability of the AC sources is

demonstrated with reasonable assurance. Also, the reliability of the

EDGs will be enhanced by reducing the potential for numerous

unloaded EDG starts during an LCO [limiting condition for operation]

period. The proposed changes could reduce the number of required

unloaded EDG starts from nine to one. Therefore, these 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 to rewrite the action statements for

Technical Specification 3.8.1.1 do not alter the method of operating

the plant, nor do they introduce any new failure modes. The proposed

changes affect EDG testing frequency only, they have no impact on

any accident analysis. The proposed changes provide assurance that

the EDGs will be able to power their respective safety systems if

required. Also, they do not involve any physical alterations to

plant equipment or procedures which would introduce any new or

unique operational modes or accident precursors.

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

The proposed changes to rewrite the action statements for

Technical Specification 3.8.1.1 do not affect the capability of the

EDGs to perform their function. The intent of the changes is to

increase the overall EDG reliability, by reducing the wear resulting

from excessive and unwarranted testing. 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: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, City Place, Hartford, Connecticut 06103-3499.

NRC Project Director: John F. Stolz

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: June 12, 1992, as supplemented September

17, 1992, March 17, 1993, August 17, 1993, August 18, 1993, and

December 29, 1993.

Description of amendment request: The purpose of the proposed

changes is to revise the Technical Specifications (TS) to permit

operation of the James A. FitzPatrick Nuclear Power Plant at an uprated

power of 2536 MWt. The licensee's engineering analyses and evaluations

confirm that the plant can be operated at an uprated power. The

increase in the rated power from 2436 MWt to 2536 MWt corresponds to a

4.8 percent increase in rated steam flow. The increase in rated power

remains below the plant design power level of 2550 MWt which was the

basis for the original plant safety evaluation.

The changes affect the operating parameters of the reactor,

operational restrictions, setpoints for safety systems, analytical

results, and test requirements. There are also administrative changes.

The changes in each of these categories are summarized as follows:

Reactor Parameters: The effect on reactor parameters is limited.

Higher power is achieved by control rod pattern adjustments to increase

reactor thermal power in a more uniform (flattened) powerdistribution

to increase steam flow without increasing core recirculation flow. This

requires an increased reactor dome pressure for adequate turbine inlet

pressure.

Operational Limits: The increased thermal power requires a change

to the limitation on operation in the high power low flow portion of

the power/flow map to limit thermal hydraulic instabilities and power

oscillations.

Setpoints: The increased reactor pressure has a direct impact on

the high pressure scram setpoint and the safety relief valve setpoint.

Additionally, the bypass for the turbine stop valve closure and control

valve fast closure scram will be changed in proportion to the increase

in thermal power.

Analysis Results: Analyses of uprated power transients and

accidents requires changes to various TS and their Bases. Operational

parameters and assumptions used in analyses were revised to reflect

their use as initial conditions. Revised radiological analyses changed

dose results. The results of the accident analyses requires revisions

to properly reflect plant capabilities.

Testing: A number of changes to testing requirements result from

power uprate. The increase to reactor pressure has a direct effect on

hydrostatic leakage testing pressure. The test pressure for High

Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling

(RCIC) pumps will be revised to reflect safety relief valve setpoints

assumed in analyses.

Administrative: Administrative changes (i.e., adding references,

revising references and correcting associated errors) will also be

made.

No changes to the Radiological Effluent Technical Specifications

were identified.

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 at a thermal power of 2536

MWt will 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 James A. FitzPatrick nuclear power plant was reviewed for

operation at a rated power of 2550 MWt at the time of its operating

license, [***]. This review was based on the original design of the

plant. Since that time, a number of safety issues of a generic and

plant specific nature as well as plant modifications have changed

the originally reviewed design.

Generic criteria, methodologies and evaluation scope required to

uprate BWRs up to 5% were prepared by General Electric and submitted

to the NRC in LTR-1 [NEDC-31897P-A, ``Generic Guidelines for General

Electric Boiling Water Reactor Power Uprate]. This was

supplemented by the submittal of generic evaluations in LTR-2 [NEDC-

31984P, ``Generic Evaluations of General Electric Boiling Water

Reactor Power Uprate'' and Supplement 1] to determine: which NRC and

industry generic communications were applicable to power uprate and

how they should be treated; analytical evaluations that could be

generically approved; bounding evaluations of components and

equipment, and; the effect of power uprate on safety margin. These

generic evaluations are supplemented by plant specific evaluations.

The Power Uprate Safety Analysis Report (PUSAR) describes the

dependence placed on References 1 and 2, the additional analyses

that were performed, the results of these additional analyses and

overall conclusions on the safety impacts of power uprate.

The plant systems and components will be within design limits at

power uprate conditions with minor modifications. At uprated power,

the power plant will not be operated in a manner that is different

from current operations except for limited changes to operating

parameters such as primary system pressure, steam flow and feedwater

temperature. Setpoints are revised as necessary to reflect new

operational conditions and analyses. The ECCS-LOCA [emergency core

cooling system-loss-of-coolant accident] analysis using current

practices demonstrates compliance with design and regulatory

acceptance criteria at uprated power.

The radiological consequences of accidents have been evaluated

using more current methodologies with consistent assumptions and

continue to meet acceptance criteria. Compliance with NRC dose

criteria using current methodologies is discussed in Section 9.2 of

the PUSAR. The effect of power uprate on dose analyses now discussed

in the FSAR [Final Safety Analysis Report] were qualitatively

assessed recognizing that power uprate increases doses in direct

proportion to the 4.1% increase in thermal power. An increase of

4.1% to the calculated doses currently identified in FSAR Chapter 14

indicates that a reevaluation using the original methodology would

have demonstrated compliance with current NRC dose criteria. A

review of Table 14.4-2 indicates that, with the 4.1% increase,

offsite doses would be substantially less than NRC allowable values.

A review of Table 14.8-1 indicates that, with the 4.1% increase,

control room doses would be substantially less than NRC allowables

except for the main steam line break (MSLB). However, the MSLB dose

would drop well below allowables once the proposed change on

allowable coolant activity (reduces the limit by more than a factor

of ten) is accounted for.

[Thus, based on the above analysis and supplemental analyses

performed by the NRC staff, the increase in power level discussed

herein and associated Technical Specification changes do not

significantly increase 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.

Operation at uprated power involves no changes to the manner in

which the plant is operated. There are changes to operational

parameters and setpoints but analyses of these identified no new

failure modes or accident scenarios. The effects of transients and

accidents fall within design capabilities. Systems and components

are capable of operating and performing their safety functions at

uprated power. No mechanisms for creating a new or different

accident were identified.

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

The power uprate will not result in significant increases to

primary system temperature and pressure due to postulated operating

transients or accidents. These and other margins of safety have been

discussed in the PUSAR, where it is demonstrated that there will be

no reductions in the margin of safety because the plant will still

meet its design and regulatory acceptance criteria. For example, the

core will continue to be operated with the same margin to the safety

limit minimum critical power ratio. Fuel thermal limits will

continue to meet NRC acceptance criteria. Plant systems and

equipment are designed for uprated power conditions and have been

evaluated for their capability to perform at uprated conditions.

They will continue to perform within design limits.

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: Robert A. Capra

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 22, 1993

Description of amendment request: The proposed amendment to the

James A. FitzPatrick Technical Specifications proposes to remove the

reference to American Society for Testing and Materials (ASTM) Standard

D 270-65 from Surveillance Requirement 4.12.A.1.i. ASTM D 270-65, which

specifies procedures to draw a representative fuel oil sample, has been

superseded and is no longer in effect. The FitzPatrick Surveillance

Procedure currently requires sampling in accordance with ASTM D 270-65

but notes that it has been superseded by ASTM D 4057-88. The proposed

change makes Surveillance Requirement 4.12.A.1.i consistent with fuel

oil Surveillance Requirement 4.9.C.1, for the emergency diesel

generators (EDGs), by adopting the current industry fuel oil sampling

standard. There are no changes to the acceptance criteria for fuel oil

quality which are based on ASTM D 975-81.

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.

Plant accident analyses are not affected by the Technical

Specification change. The change removes reference to ASTM D 270-65

as the method for obtaining samples of diesel fuel oil from

Surveillance Requirement 4.12.A.1.i. The American Society for

Testing and Materials has superseded ASTM D 270-65 with ASTM D 4057.

The proposed change provides improved flexibility to adopt standards

as they are issued without requiring a license amendment and makes

Surveillance Requirement 4.12.A.1.i consistent with both

Surveillance Requirement 4.9.C.1 for the EDG and the STS [Standard

Technical Specification]. The nature of this change will not cause

any increase in the probability or consequences of previously

evaluated accidents.

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

from those previously evaluated.

The proposed change involves no hardware modifications to any

plant structures, systems or components. The change removes

reference to ASTM D 270-65 as the method for obtaining samples of

diesel fuel oil from Surveillance Requirement 4.12.A.1.i. The

American Society for Testing and Materials has superseded ASTM D

270-65 with ASTM D 4057. The nature of this change is such that no

new or different kind of accident can be created.

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

The proposed change will not cause a reduction in the margin of

safety. The results of the plant accident analyses continue to bound

operation under the proposed changes so there is no reduction in the

margin of safety. The change removes reference to ASTM D 270-65 to

allow the use of ASTM D 4057-88 as the method for obtaining samples

of diesel fire pump fuel oil. The change will make Surveillance

Requirement 4.12.A.1.i consistent with existing Surveillance

Requirement 4.9.C.1 for the EDG and allow current standards to be

used for fuel oil sampling. Revisions of this nature will not cause

a 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: Robert A. Capra

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 28, 1993

Description of amendment request: The proposed amendment to the

James A. FitzPatrick Technical Specifications (TSs) revises the scope

of startup and power escalation reports to reflect the guidance

provided by the Standard Technical Specifications (STS). Current TS

requirements state, ``The report shall address each of the tests

identified in the FSAR [Final Safety Analysis Report] ....''

The list of tests provided in the FSAR includes many tests which

were required for the initial plant startup but are not performed for

subsequent startups. Currently, the licensee satisfies TS requirements

in the startup reports by identifying tests which are not performed,

their purpose, and the reason for not performing them.

The inclusion of such a section in the startup reports is

unnecessary. This amendment would revise the TSs to permit subsequent

startup reports to address only those tests that are actually

performed.

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 proposed changes involve no hardware modifications, no

changes to the operation of any system or component, no changes to

structures, and alters future startup reports by addressing only

tests that are performed. A change in reporting format will not

eliminate the performance of startup tests that are necessary or

required. These changes can not alter the probability or

consequences of existing accident analyses as documented in the FSAR

or the NRC staff SER [Safety Evaluation Report].

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

from those previously evaluated.

The changes do not alter the testing procedures or

methodologies. The changes provide a basis, in accordance with NRC

guidance, for allowing future startup test reports to mention and

discuss only those tests that are actually performed. These changes

can not result in a new or different type of accident than those

previously evaluated.

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

There are no changes to tests that are performed in support of

plant startup [***]. Removing a requirement for reporting on tests

that are not performed will not cause a reduction in any margin of

safety.

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 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: Robert A. Capra

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 29, 1993

Description of amendment request: The proposed amendment to the

James A. FitzPatrick Technical Specifications would eliminate an

inconsistency between the Reactor Coolant System (RSC) leakage

detection and operability requirements in Limiting Conditions for

Operation (LCO) 3.6.D.1 and 3.6.D.4.

LCO 3.6.D.1 currently requires that RCS leakage be below specified

limits when there is irradiated fuel in the reactor vessel and reactor

coolant temperature is greater than 212 deg.F. LCO 3.6.D.4 requires the

leakage monitoring systems to be operable during power operation (i.e.,

when the mode switch is in the Startup/Hot Standby position or the Run

position with the reactor critical above 1 percent rated power, as

defined per Specification 1.0.O). These two LCOs are not consistent.

The proposed revision of LCO 3.6.D.4 will take the more conservative

approach of requiring the leakage monitoring systems to be operable

when the leakage limits of LCO 3.6.D.1 are in effect.

The proposed changes also make editorial corrections which are

considered administrative in nature.

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 proposed amendment involves no hardware changes, no changes

to the operation of any systems or components and no changes to

structures. It alters an LCO to require plant leakage detection

systems to be operable during the same plant conditions that RCS

leakage limits apply. The revised LCO requires leakage detection

systems in operational modes consistent with other portions of the

Technical Specifications. Additional changes include editorial

corrections such as correct specification numbering, proper system

identification, and clarification of a surveillance requirement

cross reference. Since the change to the LCO for leakage detection

will require operability under a greater range of plant conditions

to be consistent with detection requirements, there is no change to

previously determined accident probabilities or consequences. The

editorial changes have no adverse safety implications.

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

from those previously evaluated.

The proposed amendment involves no hardware changes, no changes

to the operation of any systems or components and no changes to

structures. It alters the Technical Specifications only to the

extent of making two LCOs consistent by requiring the leakage

detection system to be operable when leakage limits apply and making

editorial changes. Editorial changes and increasing the plant

conditions for leakage system operability can not create the

possibility of a new or different kind of accident from those

previously evaluated since the editorial changes have no safety

significance and the operability changes are being made for

consistency with the modes when leakage detection is required to

function.

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

The proposed amendment revisions involve no hardware changes, no

changes to the operation of any systems and no changes to

structures. The revised LCO criteria for RCS leakage detection

system operability has increased the plant conditions when

operability is required to match the plant conditions when leakage

limits apply. Editorial changes and expanded operability

requirements in the LCOs will not result in any change to existing

safety margins.

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: Robert A. Capra

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 29, 1993

Description of amendment request: This proposed amendment to the

James A. FitzPatrick Technical Specifications, Appendix B, revises

Surveillance Requirement 3.1.a and Table 3.10-2. The existing

Surveillance Requirement in Specification 3.1.a, which references Table

3.10-2, requires the performance of instrument checks, tests, and

calibrations to assure the operability of specific gaseous effluent

radiation monitors. It does not require any surveillances for the

associated data recorders. However, Limiting Condition for Operation

(LCO) 3.1.a specifically states that ''...pathways shall be monitored

and recorded ....'' which requires the recorders to be operable for the

gaseous effluent monitoring system to be considered operable. The

proposed changes to the Radiological Effluent Technical Specifications

(RETS) add the radiation monitor recorders to the Surveillance

Requirement and Table 3.10-2 for the gaseous effluent monitoring

system. This will provide the surveillance requirements for the data

recorders.

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 proposed amendment involves no hardware changes, no changes

to the operation of any systems or components and no changes to

structures. It alters procedures by including the recorders in the

Surveillance Requirements for determining the operability of the

radiological monitoring/recording instrumentation. The inclusion of

the data recorders in the Surveillance Requirements for checking,

testing, and calibration does not reduce gaseous effluent monitoring

capability while providing greater assurance that data is recorded.

Adding the existing system surveillance to the Technical

Specifications will not have any affect on previously evaluated

accident probabilities or consequences.

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

from those previously evaluated.

The proposed amendment involves no hardware changes, no changes

to the operation of any systems or components and no changes to

structures. It alters procedures by including the data recorders in

the Surveillance Requirements for the radiation monitors. These

changes do not affect the manner in which the gaseous effluent

radiation monitoring system is operated or tested. The resulting

changes do not pose a safety issue concern different from those

analyzed previously in the FSAR [Final Safety Analysis Report] or

the NRC staff SER [Safety Evaluation Report].

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

The proposed amendment involves no hardware changes, no changes

to the operation of any systems and no changes to structures. It

alters procedures by including the data recorders in the

Surveillance Requirements for the radiation monitors. The revised

surveillance requirement increases the scope of surveillance for the

gaseous effluent radiation monitoring system, for consistency with

the LCO. It does not cause any reduction in any safety margins.

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: Robert A. Capra

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: January 11, 1994

Description of amendment request: This proposed amendment to the

James A. FitzPatrick Technical Specifications (TS) would temporarily

waive the 2-year maximum surveillance interval for the Type C test

(local leak rate test) of the shutdown cooling isolation valves (10MOV-

17 and 10MOV-18). The waiver would permit deferring this test until the

next refueling outage, currently scheduled for the end of November

1994.

TS 4.7.A.2.e(5) and 10 CFR Part 50 Appendix J require the

containment isolation valves to be leak-rate tested during each

refueling outage, but at intervals of not greater than 2 years. As a

result of the extended 1991-1993 refueling outage, and the length of

the current operating cycle, tests of all containment isolation valves

are due prior to the start of the next refueling outage. Two mid-cycle

outages were scheduled, in part, to accommodate the testing of these

valves, except for the shutdown cooling isolation valves. The shutdown

cooling isolation valves cannot be removed from service during a non-

refueling outage to accommodate a leak-rate test since its associated

system is needed to remove reactor decay heat. The reliability of the

isolation valve design, and the very low probability that the shutdown

cooling system penetration would result in a pathway for leakage to the

reactor building, as discussed in the application, is provided as

justification for this one-time schedular extension.

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 proposed change is limited to a one-time schedular extension

in the shutdown cooling isolation valve Type C test. The change does

not introduce any new modes of plant operation, make any physical

changes, or alter any operational setpoints. Therefore, the change

does not degrade the performance of any safety system assumed to

function in the accident analysis. The extension will not result in

a significant increase in valve leakage considering that: (1) the

valves are infrequently stroked, and then only when the reactor

vessel is at low pressure, (2) monitoring the normal operating

status of the RHR [residual heat removal] system assures the absence

of gross valve leakage, and (3) the valves were replaced with valves

of an improved design which has been confirmed by past Type C tests

to exhibit satisfactory leak rate performance. For these reasons,

the change does not involve a significant increase in the

probability of an accident.

The change does not involve a significant increase in the

consequences of an accident evaluated since any leakage through the

shutdown cooling penetration will not significantly increase for

reasons discussed in the previous paragraph, and such leakage is

negligible compared to the main steam line break accident analyzed

in the FSAR [final safety analysis report].

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

from those previously evaluated.

The proposed change does not introduce new accident initiators

or failure mechanisms since the change does not alter the physical

characteristics of any plant system or component. The change is

limited to a one-time schedule extension for the shutdown cooling

isolation valve Type C tests.

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

There is a very low probability of a significant increase in

valve leakage considering the demonstrated reliability of the

current valve design, the infrequent use of the valves, and the

monitoring of the normal operating status of the RHR system.

Moreover, any potential incremental benefit of performing the tests

within the two year requirement would not be sufficient to offset

the increased occupational radiation exposure associated with

testing, and the risk to plant safety associated with the removal

from service of the primary method of decay heat removal.

Consequently, 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: 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: Robert A. Capra

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

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

Date of amendment request: October 1, 1993 TS 93-09

Description of amendment request: The proposed changes would revise

the setpoints and time delays for the auxiliary feedwater (AFW) and 6.9

kv shutdown board loss of voltage and degraded voltage instrumentation.

The proposed changes would affect Technical Specification (TS) Tables

3.3-3, 3.3-4, 3.3-5, 4.3-2, and the Bases for Specification 3/4.7.1.2.

Table 3.3-3 would be revised: (1) to reflect the use of a two-out-of-

three voltage sensor logic for loss of power detection (and one-out-of-

two logic scheme for the timing relays) by adding requirements for the

respective AFW, 6.9 kv shutdown board, and emergency diesel generator

(EDG) voltage sensors and load shed timers; (2) by changing the

description of the functional unit from ``Station Blackout'' to ``Loss

of Power Start;'' (3) by adding a footnote to indicate that the new

requirements apply only to the shutdown boards on the same unit; (4) by

increasing the modes for which the EDG sensors and timers must be

operable to include Modes 5 and 6 when the associated EDG must be

operable; (5) by changing the associated Action statements to reflect

the design changes and consistency; (6) to reflect consistent

terminology in the Table and action statements; (7) by changing the

minimum number of channels that would be required to be operable; (8)

by changing the footnote to reflect the conditions when the loss-of-

power instrumentation is required to be operable in Modes 5 and 6; and

(9) by adding an exclusion to Specification 3.0.4. Similar changes were

proposed for Tables 3.3-4, 3.3-5 and 4.3-2. The proposed change to

Bases 3/4.7.1.2 would clarify AFW operability on loss of power.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed revision supports the implementation of design

logic and setpoint changes to the loss-of-power relaying. This

relaying is designed to ensure adequate voltage is available to

safety-related loads in order to enhance their operability and

support accident mitigation functions and to provide for auxiliary

feedwater (AFW) pump starts. The design changes alter relay logic

and delete unnecessary relaying, but do not change the diesel

generator (D/G) start and load-shedding actuations that result from

loss-of-power conditions. Therefore, no new actuations or functions

have been created; and because the existing and proposed functions

provide for accident mitigation considerations that are not the

source of an accident, the probability of an accident is not

increased. The deletion of the 6.9-kilovolt shutdown board normal-

feeder undervoltage relays actually reduces the potential for

inadvertent shutdown board blackouts as a result of short-duration

voltage transients or instrument failures.

The setpoints and time delays for loss-of-power functions have

been modified based on the guidelines developed by the Electrical

Distribution System Clearinghouse as evaluated and determined

through detailed analysis by TVA. This design is documented in TVA

Calculations SQN-EEB-MS-TI06-0008, 27DAT, and DS-1-2 and is

available for NRC review at the SQN site. The assigned values are

conservative settings that will ensure adequate voltage is supplied

to safety-related loads for accident mitigation and safety functions

under normal, degraded, and loss-of-offsite-power voltage conditions

with appropriate time delays to prevent damage to electrical loads

and minimize premature or unnecessary actuations. The identification

of loss-of-voltage conditions is enhanced by the design changes to

ensure the timely sequencing of loads onto the D/G and the

initiation of AFW pump starts for accident mitigation. Because there

are no reductions in safety functions resulting from the design

logic, setpoint, and time-delay changes to the loss-of-power

instrumentation and offsite dose levels for postulated accidents

will not be increased, the consequences of an accident are not

increased.

The applicable mode addition, TS 3.0.4 exclusion deletion, and

response time measurement clarification incorporated in the proposed

change do not affect plant functions. These changes reflect the

requirements that SQN has been maintaining and serve to clarify the

requirements to provide consistency of application and easier

understanding. The AFW footnote addition and bases revision only

clarify operability conditions that are consistent with the plant

design for the AFW pump and loss-of-power instrumentation. Because

there are no changes to plant functions or operations, these

revisions have no impact on accident probabilities or consequences.

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

from any previously analyzed.

As described above, the loss-of-power instrumentation ensures

adequate voltage to safety-related loads by initiating D/G starts

and load shedding and provides for AFW pump starting, but is not

considered to be the source of an accident. Although the design

logic, setpoint, and time-delay actuation criteria have changed, the

output functions to various plant systems that actuate for load

shedding and D/G starts remain the same. Therefore, actuation

criteria have been affected, but not safety functions, and the TVA

evaluation has confirmed that the new design enhances the ability to

maintain adequate voltage to support safety functions. Since safety

functions have not changed and the new loss-of-power instrumentation

design continues to support operability of safety-related equipment,

no new or different accident is created.

The applicable mode addition, TS 3.0.4 exclusion deletion, and

response time measurement clarification, as well as the AFW

operability clarifications, do not affect plant functions and will

not create a new accident.

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

The proposed loss-of-power TS changes support design logic,

setpoint, and time-delay requirements that have been verified by TVA

analysis to provide acceptable voltage levels for safety-related

components. In determining the acceptability of these voltage

levels, the minimum voltage for operation as well as detrimental

component heating resulting from sustained degraded-voltage

conditions were considered. This design ensures that safety-related

loads will be available and operable for normal and accident plant

conditions. The applicable mode addition, TS 3.0.4 exclusion

deletion, response time measurement clarification, and AFW

operability clarifications provide enhancements to TS requirements

and do not affect plant functions. Therefore, no safety functions

are reduced by these changes and there is no reduction in the margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: November 18, 1993 (TS 93-17)

Description of amendment request: The proposed change would delete

the requirement of License Condition 2.H of the Operating License. This

license condition requires reporting the violation of certain license

conditions to the NRC Regional Administrator within 24 hours by phone

and facsimile, and a followup report within 14 days.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change is a deletion of an administrative reporting

requirement that does not in any way affect a previously analyzed

accident.

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

from any previously analyzed.

Since there are no changes in the way the plant is operated, the

potential for an unanalyzed accident is not created. The proposed

change is administrative in nature and does not affect any accident

initiators for SQN. No new failure modes are introduced.

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

Plant safety margins are established through limiting conditions

of operation, limiting safety system settings, and safety limits

specified in the TSs. As a result of the proposed amendment, there

will be no changes to either the physical design of the plant or to

any of these settings and limits. The proposed changes are

administrative and do not affect the safe operation of SQN.

Therefore, there will be no changes to any of the margins of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: December 7, 1993 (TS 93-16)

Description of amendment request: The proposed change would

incorporate the following changes to Section 6.0, ``Administrative

Controls,'' of the Technical Specifications (TS): (1) delete Section

6.1.2 that indicates that the Corporate Manager of Radiological Control

has the responsibility for the radiological environmental program, dose

calculations, and projections; (2) added a requirement in Section

6.5.1.6 for review of the Offsite Dose Calculations Manual (ODCM) by

the Plant Operations Review Committee (PORC); (3) added implementation

of the ODCM to Section 6.8.1 as an activity requiring written

procedures and deleted the requirement that they be maintained by the

Radiological Control Group; (4) delete the requirement in Section

6.8.4.a for Radiological Control to implement and control the ODCM; (5)

changed the review authority for changes to the ODCM from the

Radiological Assessment Review Committee (RARC) to the PORC and added a

reference to Specification 6.5.1A in Section 6.14.1.2; (6) approval

authority for deviations from the overtime guidelines in Section

6.2.2.g would be changed to show that Plant Manager designee also has

the authority; (7) PORC member titles would be changed in accordance

with the current organizational structure; (8) the requirements for the

RARC would be deleted from Sections 6.5.2.7.i, 6.5.3, 6.10.2.k and the

index; (9) move the condenser inleakage monitoring requirement from

Section 6.8.5.c.(vii) to 6.8.5.c.(iii) and delete the requirement to

repair, plug, or isolate leaks; (10) change the title ``Shift

Supervisor'' (SS) to ``Shift Operations Supervisor'' (SOS) in various

locations in Section 6 and in Operating License Items 2.C.(23).A and

2.C.(16).a for Units 1 and 2 respectively; (11) change the title of the

Senior Vice President, Nuclear Group, in Section 6.2.1 to Senior Vice

President, Nuclear Power; (12) change the title of the Operational

Quality Assurance Program in Section 6.5.2.8.d to Nuclear Quality

Assurance Program; and (13) move the requirement for implementation of

the Quality Assurance Program for environmental monitoring from Section

6.8.4.b to Section 6.8.1.h. Other administrative changes related to

these changes were also submitted that affect Section 6.0, the

operating license, index, and definitions.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). The operation of Sequoyah Nuclear Plant (SQN) in

accordance with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes only affect the administrative controls

found in Section 6.0 of the SQN TSs and the operating license. No

plant equipment or operating practices are affected by these

changes. The revised administrative controls will continue to

adequately implement administrative activities to support plant

nuclear safety. Since there are no physical changes to the plant,

there is no increase in the probability of an accident because these

administrative controls are not the source of previously evaluated

accidents. Similarly, with no change to plant equipment or operating

requirements, the plant response to accident conditions and

therefore the consequences of an accident remain unchanged. These

proposed changes will not increase the consequences of an accident

and offsite dose rates will not be impacted.

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

from any previously analyzed.

The administrative controls affected by the proposed changes are

not considered to be the source of any accident and these changes

will not alter any plant features or processes. Therefore, the

proposed changes will not create the possibility of a new or

different kind of accident and the administrative controls will

continue to implement the actions necessary to support plant

activities and nuclear safety.

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

Plant features and setpoints remain unchanged by the proposed

changes to the administrative controls. The margins of safety

established by the SQN design are not affected by these changes. The

proposed administrative controls will continue to maintain the

actions and programs that ensure appropriate plant design,

operation, and procedures to support the required margin of safety.

Therefore, the proposed changes will not reduce the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library,1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Manitowoc County, Wisconsin

Date of amendment request: September 29, 1993

Description of amendment request: The proposed amendments would

change the test intervals from monthly to quarterly, consistent with

the Inservice Test Program, for several pumps and related systems,

including safety injection, residual heat removal and containment spray

pumps.

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 amendments will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The addition of specific general considerations related to

equipment surveillance requirements and their relationship to

equipment operability are administrative in nature. They do not

change the interpretation or intent of the Technical Specifications.

These conditions are consistent with the Westinghouse Standard

Technical Specifications (STS).

The addition of the specific requirement for the Inservice Test

Program to Section 15.4.2 of the Technical Specifications only

reiterates the requirements of 10 CFR 50.55a(g). This requirement

does not implement any new requirements on the operation or testing

of equipment.

The decrease in the number of equipment operational transients

due to the increase in the surveillance interval for the Safety

Injection System (SI), Residual Heat Removal (RHR) System, and

Containment Spray (CS) System pumps and valves will result in an

increase in system availability. Reduced testing is also expected to

have a positive affect on overall equipment reliability since

frequent testing results in increased wear and potential for

equipment failure. Other actions including a monthly verification of

system lineups for the SI, CS and RHR systems provides increased

assurance of system operability between surveillance tests. The

potential for equipment problems to go undetected for a longer

period of time is small as indicated by equipment surveillance

history.

Therefore, these changes will not effect the probability or

consequences of previously analyzed accidents.

2. The proposed amendments will not create the possibility of a

new or different kind of accident from any accident previously

analyzed.

These changes only affect the equipment testing frequency.

Equipment design, operation and the methods of testing will not be

changed. Therefore, the proposed changes cannot create the

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

previously evaluated.

3. The proposed amendments will not involve a significant

reduction in the margin of safety.

The proposed changes which implement the requirements of 10 CFR

50.55a(g) and clarify the general considerations related to

equipment surveillances are administrative in nature. They do not

change the intent of any existing license or other requirement.

The increase in equipment surveillance intervals will result in

an improvement in equipment and system availability and reliability.

Surveillance of equipment will be performed as required by the

regulations and Section XI of the ASME Boiler and Pressure Vessel

Code. These proposed changes will reduce the potential for equipment

failures due to unnecessary testing.

Adequate assurance is provided by testing in accordance with the

ASME Code requirements and periodic verification of system lineups

to ensure that the affected systems remain operable and capable of

performing their design function. Therefore, a reduction in a margin

of safety will not occur.

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

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: John N. Hannon

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: December 1, 1993

Description of amendment request: The proposed amendment would

revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications

(TS) by incorporating technical and administrative changes to TS 3.10,

Control Rod and Power Distribution Limits. The proposed changes

eliminate specifications for fuel designs no longer used at Kewaunee,

specify required actions to be taken upon exceeding control bank

insertion limits, and revise the limits for Departure from Nucleate

Boiling (DNB) related parameters to assure operation within the

assumptions of the Updated Safety Analysis Report (USAR) analyses.

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) TS 3.10.b.1, 3.10.b.4 and Table TS 3.10-2

The proposed changes were reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes eliminate the specifications for fuel that

is no longer used at the Kewaunee Nuclear Power Plant. Eliminating

these specifications does not affect the probability of any accident

previously evaluated.

The specifications for the current fuel vendor are being

retained and ensure the consequences of previously evaluated

Departure from Nucleate Boiling (DNB) related accidents are

enveloped by the Updated Safety Analysis Report (USAR) analyses.

Therefore, these changes will not increase the consequences of an

accident previously evaluated in the USAR.

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

from any accident previously evaluated.

The proposed change does not alter the plant configuration,

operating setpoints, or overall plant performance. Therefore, it

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

accident.

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

The proposed change deletes specifications for fuel that is no

longer used at the Kewaunee Nuclear Power Plant. The limits for the

current fuel vendor are retained and are not affected by this

proposed change. This does not alter the input or assumptions of the

safety analysis, and therefore it will not involve a reduction in

the margin of safety.

(b) TS 3.10.d

The proposed change was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed change will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change specifies the required actions to be taken

when the control bank insertion limits are exceeded. The current TS

requires compliance with the control bank insertion limits, but

gives no corrective action for when these limits are exceeded. The

new specification requires operators to initiate boration to restore

shutdown margin within one hour of exceeding the control bank

insertion limits and to restore the control banks to within the

limits within 2 hours. If either of these requirements cannot be

achieved, within 1 hour the operators must initiate actions to

achieve Hot Standby within the next 6 hours and Hot Shutdown within

the following 6 hours. Adding these requirements clarifies and

enhances the Technical Specifications and will have no impact on the

probability of an accident previously evaluated.

The proposed addition is conservative and ensures that proper

and adequate measures are taken when the control bank exceeds the

control bank insertion limits. Therefore, this addition will not

increase the consequences of an accident previously evaluated in the

USAR.

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

from any accident previously evaluated.

The proposed change does not alter the plant configuration,

operating setpoints, or overall plant performance. Therefore, it

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

accident.

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

The proposed change is conservative and clarifies the necessary

actions to be taken when control bank insertion limits are exceeded.

This proposed change is an enhancement to the specification and does

not reduce the margin of safety.

(c) TS 3.10.k

The proposed change was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed change will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change decreases the maximum RCS inlet temperature

limit for steady state 100% operation from 536.5 degrees Fahrenheit

to 535.5 degrees Fahrenheit. The value of 539.5 degrees Fahrenheit

is the assumed RCS inlet temperature for the DNB related accidents

analyzed in the USAR. These accidents are the Uncontrolled Rod

Cluster Control Assembly Withdrawal at Power Accident, the

Uncontrolled Rod Cluster Control Assembly Withdrawal from a

Subcritical Condition Accident, the Rod Cluster Control Assembly

Misalignment, the Start-Up of an Inactive Reactor Coolant Loop

Accident, the Excessive Heat Removal due to Feedwater System

Malfunction Accident, the Excessive Load Increase Incident, the Loss

of Reactor Coolant Flow Accident, the Loss of External Electrical

Load, the Steam Line Break, and the Rod Cluster Control Assembly

Ejection. A four degree assumed instrument error reduces the maximum

allowed RCS inlet temperature to 535.5 degrees Fahrenheit.

Decreasing this value to ensure consistency with the USAR analysis

assumptions will have no impact on the probability of an accident

previously evaluated.

The proposed change is conservative to ensure that the

consequences of a previously evaluated DNB-related accident are

enveloped by the USAR analysis. Therefore, this change will not

increase the consequences of an accident previously evaluated in the

USAR.

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

from any accident previously evaluated.

The proposed change does not alter the plant configuration,

operating setpoints, or overall plant performance. Therefore, it

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

accident.

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

The proposed change is conservative and is consistent with the

assumptions in the USAR. This proposed change is an enhancement to

the specification and does not reduce the margin of safety.

(d) TS 3.10.l

The proposed change was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed change will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change increased the minimum RCS pressure limit for

steady-state 100% power operation from the currently specified 2200

psig to 2205 psig. The value of 2205 psig (30 psig below the nominal

design value of 2235 psig) was the assumed initial condition for the

DNB related accidents analyzed in the USAR. These accidents are the

Uncontrolled Rod Cluster Control Assembly Withdrawal at Power

Accident, the Uncontrolled Rod Cluster Control Assembly Withdrawal

from a Subcritical Condition Accident, the Rod Cluster Control

Assembly Misalignment, the Start-Up of an Inactive Reactor Coolant

Loop Accident, the Excessive Heat Removal due to Feedwater System

Malfunction Accident, the Excessive Load Increase Incident, the Loss

of Reactor Coolant Flow Accident, the Loss of External Electrical

Load, the Steam Line Break, and the Rod Cluster Control Assembly

Ejection. Increasing this value to ensure consistency with the USAR

analysis assumptions will have no impact on the probability of an

accident previously evaluated.

The proposed change is conservative to ensure that the

consequences of a previously evaluated DNB-related accident is

enveloped by the USAR analysis. Therefore, this change will not

increase the consequences of an accident previously evaluated in the

USAR.

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

from any accident previously evaluated.

The proposed change does not alter the plant configuration,

operating setpoints, or overall plant performance. Therefore, it

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

accident.

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

The proposed change is conservative and is consistent with the

assumptions in the USAR. This proposed change is an enhancement to

the specification and does not reduce the margin of safety.

(e) TS 3.10.m

The proposed change was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed change will not:

1)Involve a significant increase in the probability or

consequences of an accident previously evaluated.

TS 3.10.m.1 provides the limits and required actions to be taken

when the RCS flow rate per loop is less than the USAR analysis

assumptions. Decreasing the flow limit to be consistent with the

USAR assumptions will have no impact on the probability of an

accident previously evaluated. Compliance with the flow limit

assumed in the USAR analyses ensures the consequences of previously

evaluated DNB-related accidents are enveloped by the USAR analyses.

Therefore, this change will not increase the consequences of an

accident previously evaluated in the USAR.

The proposed revision places an additional restriction on RCS

flow when less than the design flow rate of 89,000 gallons per

minute per loop. The intent of specifying action in accordance with

3.10.n is to ensure that reactor power is reduced to a point at

which the DNB ratio margin is restored. Compliance with this

specification will not increase the probability of an accident

previously evaluated, nor increase the consequences of an accident.

The intent of TS 3.10.m.2 is to clarify the conditions under

which the reactor coolant flow rate is verified. The conditions,

methodology, and uncertainties associated with this verification are

not changed by this specification. Clarifying the TS by inclusion of

the conditions for verifying the flow rate will not increase the

probability or consequences of any accident previously evaluated.

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

from any accidents previously evaluated.

A new or different kind of accident from those previously

evaluated will not be created by this TS change. The proposed

amendment does not alter the plant configuration, operating

setpoints or overall plant performance.

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

A specification on steady-state RCS flow rate is necessary to

ensure DNB ratio criteria will be met during the DNBR limiting

events analyzed in the USAR. Reducing the reactor coolant flow limit

to the value assumed in the USAR analyses does not result in a

reduction in the margin of safety.

The additional restrictions being imposed if the specified limit

is not met provide additional assurance the DNBR margin will be

restored. These additional restrictions do not exist in the current

TS. The imposition of these restrictions results in an enhancement

to the margin of safety.

Clarifying the TS by inclusion of conditions under which the

flow verification is to be performed will not reduce the margin of

safety. Existing approved constraints, methodology, and

uncertainties are not being changed by this clarification.

(f) TS 3.10.n

The proposed change was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed change will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The intent of this new TS is to outline the actions required

when the limits of TS 3.10.k (RCS temperature), TS 3.10.l (RCS

pressure) and TS 3.10.m.1 (RCS flow) are exceeded. Collectively,

these three specifications place limits on the DNB-related

parameters to assure each is maintained within the normal steady-

state envelope assumed in the USAR safety analysis. This

specification is an enhancement to our existing specification to add

clear guidance which does not presently exist. Providing this

information for the plant staff and operators will not increase the

probability of an accident previously evaluated.

The addition of this action statement will ensure that the

consequences of an analyzed accident are not increased. The proposed

specification allows 2 hours to evaluate and restore parameters to

within limits. If this time frame is not satisfied, then within the

next 6 hours, power is reduced in order to restore a margin of

safety. Following analysis, thermal power may be raised not to

exceed a level analyzed to maintain a minimum DNBR of 1.30.

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

from any accident previously evaluated.

The proposed amendment does not alter the plant configuration,

operating setpoints or overall plant performance.

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

Addition of the specification is an enhancement to the current

specification which does not alter input to the safety analysis.

Therefore, it will not involve a reduction in the margin of safety.

(g) Administrative changes to Section TS 3.10 including Figure

TS 3.10-2.

The proposed change was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will 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 the margin of safety.

The proposed changes are administrative in nature and do not

alter the intent or interpretation of the TS. Therefore, no

significant hazards exist.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Wisconsin

Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin

54301.

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

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

NRC Project Director: John N. Hannon

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: December 7, 1993

Description of amendment request: The proposed amendment would

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

5.3.a.1 to provide flexibility in the repair of fuel assemblies

containing damaged and leaking fuel rods by reconstituting the

assemblies, provided that an NRC-approved methodology is used. This

proposed change is consistent with guidance provided in Supplement 1 to

Generic Letter (GL) 90-02, ``Alternative Requirements for Fuel

Assemblies in the Design Features Section of Technical

Specifications,'' dated July 21, 1992. In addition, administrative

changes to KNPP TS Section 5 have been proposed.

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 proposed changes were revised in accordance with the

provision of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will not:

1) involve a significant increase in the probability or

consequences of an accident previously evaluated.

This proposed change to the requirements for ``Fuel Assemblies''

in the ``Design Features'' section of the KNPP TS will not involve a

significant increase in the probability or consequences of an

accident previously evaluated. This proposed change will not result

in modifications to fuel assemblies that would have a significant

effect on safety because of the requirement to implement these

changes using an NRC-approved methodology. This requirement will

confirm conformance to existing design limits and confirm that

safety analyses criteria are met before operation during the next

fuel cycle. This license amendment request is consistent with

guidance provided by the NRC and will result in flexibility for

improved fuel performance.

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

from any accident previously evaluated.

The creation of new or different kind of accident from any

previously evaluated accident is not considered a possibility

because the changes are administrative in nature and do not

represent an actual modification to the plant or change its safety

analyses.

3) involve a significant reduction in the margin of safety.

The margin of safety is maintained by adherence to other fuel

related TS limits and the USAR design bases. The changes do not

directly affect any safety system or the safety limits, and thus

does not affect the plant margin of safety.

Accordingly, these proposed changes do not involve a significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Wisconsin

Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin

54301.

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

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

NRC Project Director: John N. Hannon

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 20555, and at the local public document

rooms for the particular facilities involved.

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

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

County, North Carolina

Date of application for amendments: October 19, 1993

Brief description of amendments: The amendments change the

Technical Specifications (TS) add a footnote to TS 4.6.1.2.b that

allows a one-time exemption from the accelerated containment integrated

leak rate test (CILRT) requirements to return the CILRT frequency for

both units to a normal Type A test interval.

Date of issuance: January 11, 1994

Effective date: January 11, 1994

Amendment Nos.: 167 and 198

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

revise the Technical Specifications.

Date of initial notice in Federal Register: November 10, 1993 (58

FR 59745) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 11, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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: May 3, 1993, as supplemented

August 11, 1993.

Brief description of amendments: The amendments revise the limiting

conditions for operation and surveillance requirements related to the

Low Pressure Service Water System.

Date of issuance: January 13, 1994

Effective date: To be implemented within 30 days from the date of

issuance.

Amendment Nos.: 203, 203, and 200

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 13, 1993 (58 FR

52983) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 13, 1994.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: October 27, 1993

Brief description of amendment: The amendment relocated the

requirement of Technical Specification 4.5.2.g.1 to verify the correct

position of each electrical and/or mechanical position stop for the

Emergency Core Cooling System throttle valves within 4 hours of each

valve stroking operation or maintenance on the valve, to procedures

that control the maintenance and operation of these valves.

Date of issuance: January 14, 1994

Effective date: To be implemented within 30 days from the date of

issuance.

Amendment No.: 155

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

Technical Specifications.

Date of initial notice in Federal Register: December 8, 1993 (58 FR

64606) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 14, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Mississippi Power &

Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit

1, Claiborne County, Mississippi

Date of application for amendment: August 11, 1993

Brief description of amendment: The amendment changed the Grand

Gulf Nuclear Station Technical Specifications to support compliance

with the new requirements of Title 10 Code of Federal Regulations Part

20 and Part 50.36a. The request to change the wording of TS 1.46 which

relates to the definition of an UNRESTRICTED AREA remains under

consideration and will be the subject of a future licensing action.

Date of issuance: January 10, 1994

Effective date: January 10, 1994

Amendment No: 111

Facility Operating License No. NPF-29. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: September 1, 1993 (58

FR 46233) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 10, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: Judge George W. Armstrong

Library, Post Office Box 1406, S. Commerce at Washington, Natchez,

Mississippi 39120.

North Atlantic Energy Service Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: August 27, 1993.

Description of amendment request: The amendment changes the

footnote on page 1 of License NPF-86 by deleting Vermont Electric

Generation and Transmission Cooperative, Inc., (Vermont), as one of the

entities for which North Atlantic Energy Service Corporation (North

Atlantic) is authorized to act. The change reflects the purchase of

Vermont's share of the Seabrook Station, Unit 1 by North Atlantic

Energy Corporation (NAEC) pursuant to a prior settlement of a claim by

Vermont against Public Service Company of New Hampshire (PSNH). NAEC

acquired PSNH's interest in the Seabrook Station, Unit 1 in accordance

with the Plan for Reorganization for PSNH.

Date of issuance: January 7, 1994

Effective date: To be implemented by May 30, 1994.

Amendment No.: 28

Facility Operating License No. NPF-86. Amendment revised the

License.

Date of initial notice in Federal Register: October 13, 1993 (58 FR

52990). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 7, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Exeter Public Library, 47

Front Street, Exeter, New Hampshire 03833.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: October 7, 1992 as supplemented July 12,

1993

Brief description of amendment: The amendment changed the setpoint

limit for the degraded-voltage protection system referred to as the

offsite-power low signal.

Date of issuance: January 14, 1994

Effective date: January 14, 1994

Amendment No.: 159

Facility Operating License No. DPR-40. Amendment revised the

TechnicalSpecifications.

Date of initial notice in Federal Register: November 25, 1992 (57

FR 55584) The additional information contained in the supplemental

letter dated July 12, 1993, was clarifying in nature and, thus, within

the scope of the initial notice and did not affect the staff's proposed

no significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated January 14, 1994. No significant hazards consideration comments

received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Omaha Public Power District, Docket No. 50-285, Fort Calhoun

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: September 15, 1993

Brief description of amendment: The amendment revised the Technical

Specifications (TSs) by implementing Generic Letters (GLs) 86-10 and

88-12. This removed the fire protection TSs and placed these parts in

the Updated Safety Analysis Report (USAR).

Date of issuance: January 14, 1994

Effective date: January 14, 1994

Amendment No.: 160

Facility Operating License No. DPR-40. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 10, 1993 (58

FR 59753) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 14, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Philadelphia Electric Company, Docket No. 50-352, Limerick

Generating Station, Unit 1, Montgomery County, Pennsylvania.

Date of application for amendment: August 3, 1993

Brief description of amendment: This amendment removes shutdown

system control valves and primary containment isolation valves from TS

Tables 3.3.7.4-1, ``Remote Shutdown Instrumentation and Controls,'' and

3.6.3-1, ``Primary Containment Isolation Valves,'' as a result of

eliminating the steam condensing mode of the Residual Heat Removal

system.

Date of issuance: January 12, 1994

Effective date: January 12, 1994

Amendment No. 65

Facility Operating License No. NPF-39. This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 29, 1993 (58

FR 50969) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 12, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

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

Date of application for amendment: May 21, 1993, as supplemented

October 7, 1993, and December 3, 1993

Brief description of amendment: The amendment revised the Technical

Specifications (TSs) to incorporate the following changes:

(1) The safety injection system test frequency (specified in TS

Section 4.5.A.1.a) was changed to accommodate operation on a 24-month

cycle.

(2) The loss of normal AC in conjunction with a safety injection

signal test frequency (specified in TS Section 4.6.A.3) was changed to

accommodate operation on a 24-month cycle. This TS section was also

reformatted to improve clarity.

(3) The auxiliary feedwater system undervoltage automatic start

test f

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