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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving no Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from April 2, 1994, through April 15, 1994. The

last biweekly notice was published on April 13, 1994 (59 FR 17591).

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 May 27, 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.

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

529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.

1, 2, and 3, Maricopa County, Arizona

Date of Amendment Requests: January 4, 1994.

Description of Amendment Requests: The proposed amendment would

change Technical Specification 3/4.2.3 Azimuthal Power Tilt and its

associated bases. The licensee proposed to change the Azimuthal Power

Tilt limit from less than or equal to 10 percent to less than or equal

to 3 percent when the Core Operating Limit Supervisory System is out of

service.

Basis for Proposed No Significant Hazards Consideration

Determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1--Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Decreasing the COLSS [Core Operating Limit Supervisory System]

out-of service Azimuthal Power Tilt Technical Specification limit

does not increase the probability or consequences of an accident

previously evaluated. The Technical Specification operating limit is

being conservatively reduced to conform to the assumptions used in

the safety analysis. The reduced operating limit requires a more

uniform power distribution in the reactor core. The uniform power

distribution may reduce the consequences of an accident previously

evaluated by not allowing regions in the core to operate at higher

power levels.

Standard 2--Create the possibility of a new or different kind of

accident from any accident previously analyzed.

The proposed amendment will result in an alarm setpoint change,

but does not involve any equipment changes and will not alter the

manner in which the plant will be operated. For this reason, this

amendment will not create the possibility of an new or different

kind of accident from any previously evaluated. The proposed

operating range is smaller and completely within the existing

Technical Specification limits; thus, there are no mechanisms to

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

those previously evaluated.

Standard 3--Involve a significant reduction in a margin of

safety.

The proposed amendment conservatively reduces the COLSS out-of-

service Azimuthal Power Tilt Technical Specification limit, thereby

increasing the margin of safety. The proposed operating range is

smaller and completely bounded by the existing Technical

Specification limits.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room Location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

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

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

Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: Theodore R. Quay.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North

Carolina

Date of Amendment Request: March 25, 1994.

Description of Amendment Request: The amendment would revise

Technical Specification 3/4.8.4.2, Motor Operated Valves Thermal

Overload Protection, with a more accurate description of the motor-

operated valve (MOV) bypass configuration.

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 is administrative in nature, providing a more

accurate description of the MOV electrical supply configuration

related to the thermal overload bypass function. Therefore, the

change in terminology would not increase 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.

The proposed change does not involve any modifications or

additions to plant equipment and the design and operation of the

plant will not be affected. Therefore, the change in MOV thermal

overload bypass function terminology would not increase 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.

The proposed terminology change does not affect any parameters

which relate to the margin of safety as defined in the Technical

Specifications or in the FSAR [Final Safety Analysis Report].

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

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

NRC Project Director: William H. Bateman.

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

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

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

County, Illinois

Date of Application for Amendments: March 7, 1994, as supplemented

on March 24, 1994.

Description of Amendment Requests: The proposed amendment would

change Technical Specification 4.6.1.2 by removing the specific

schedules for containment integrated leak rate testing (CILRT) and

specifying that the testing will be done in accordance with Appendix J

to 10 CFR part 50.

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 changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed change will allow flexibility in the scheduling for

Type A tests in the 10-year service period while still meeting the

requirements in 10 CFR 50 Appendix J. Additional flexibility is

needed for plants using an 18-month fuel cycle to allow refueling

outages and 10-year inservice testing intervals to coincide. For

performance of the third Type A test at Byron, the change would

allow an extension of four (4) months beyond the current maximum 50-

month surveillance interval. The third test would be completed at

the fifty-four (54) month interval for Byron Units 1 and 2.

For Braidwood Units 1 and 2, an extension on the surveillance

time interval will not be necessary to satisfy the requirements of

Appendix J. The Braidwood Units have scheduled the third Type A test

to be conducted with the 10-year Inservice Inspection.

The results of the previous Type A leak tests show the overall

leakage from the Byron containment buildings at very low levels. The

extension of the Type A test by four months would not cause the

consequences of a previously evaluated accident to increase. By

continuing to conform to the requirements of 10 CFR 50 Appendix J,

the test frequency, methodology, and acceptance criteria for

containment leakage remains the same. Therefore, there is no

significant increase in the probability or the consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not affect the design or operation of

any system, structure or component in the plant. There are no

changes to parameters governing plant operation and no new or

different type of equipment will be installed. No new accident

scenarios are created by the proposed change because the test

frequency continues to meet the requirements of Appendix J of 10 CFR

part 50. There is no affect on containment structure, the

penetrations, or the facility. The proposed change to the test

schedule only provides flexibility in meeting the same requirement

for three tests in a 10-year period. The testing method and bases

have not changed. Therefore, operation of the units with this more

flexible test schedule will not result in an accident previously not

analyzed in the Updated Final Safety Analysis Report (UFSAR). The

proposed changes do not impact the design bases of the containment

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

basis accident. Therefore, the changes do not create the possibility

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

evaluated.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes do not affect the margin of safety for any

Technical Specifications. The initial conditions and methodologies

used in the accident analyses remain unchanged, therefore, the

results of the accident analyses are not impacted. The proposed

change to the schedule allows for additional flexibility in meeting

the requirement for three tests in a 10-year period. Elimination of

the specified time interval for Type A testing would allow Byron

Units 1 and 2 to extend the surveillance requirement of the third

Type A test by four (4) months. This would exceed the existing

maximum 50 month interval currently specified in Technical

Specifications. The extension will allow performance of the Type A

test to coincide with the seventh refueling outage, 10 year

Inservice Inspection, and continue to meet the requirements of

Appendix J to 10 CFR part 50. These proposed changes do not affect

or change any limiting conditions for operation (LCO), or any other

surveillance requirements in the Technical Specifications.

The results of the previous Type A leak tests have shown that

the overall leakage rates from the Byron containment buildings were

at low levels. The latest test results for Units 1 and 2 were 0.0175

weight percent per day and 0.0376 weight percent per day,

respectively. The overall containment leakage rates have

consistently remained well below the acceptance criteria for Byron

Station Type A tests of 0.075 weight percent per day. The testing

method, acceptance criteria, and bases for the surveillance

requirement will not be changed by the proposed amendment.

The present test performance margins, coupled with the Type B &

C test program for monitoring and repairing individual leakage

components provides justification for the proposed change. The Type

B & C tests provide added assurance that the overall containment

integrated leakage rates remain satisfactory. No significant leakage

trends have been identified which threaten the overall containment

leakage specifications.

In summary, Commonwealth Edison concludes that this change does

not involve a significant reduction in a margin of safety because

the containment integritiy will be maintained. Testing in accordance

with Appendix J requirements ensures confidence is containment

intergity. The proposed Technical Specifications amendment will

continue to require testing that is consistent with Appendix J

requirements. Additionally, results from previous tests have shown

acceptable low overall containment leakage rates. Extension of Type

A testing for four months would not involve a signficant 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: 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.

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

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

of Amendment Request: February 18, 1994

Description of Amendment Request: This amendment is an additional

followup to the amendment request of May 29, 1992, published in the

Federal Register on July 8, 1992 (57 FR 30242), which changed the

Technical Specifications Sec. 1.0, Definitions, to accommodate a 24-

month fuel cycle and which proposed the extension of the test intervals

for specific surveillance tests. This amendment proposes extending the

surveillance intervals to 24 months for the following additional

surveillance tests:

(1) Analog Rod Position Indication.

(2) Plant Noble Gas Activity Monitor (R-44).

(3) Low Turbine Auto Stop Oil Pressure Reactor Trip.

(4) 6.9 KV Undervoltage Relays.

(5) Boric Acid Tank Level.

(6) Vapor Containment Sump Discharge Flow and Temperature

Channel.

(7) Loss of Power Undervoltage and Degraded Voltage Relays.

(8) Over-pressurization Protection System (OPS) and Control Rod

Protection System (for use with Low Parasite [LOPAR] fuel) Trip.

(9) Condenser Evacuation System Activity Monitor (R-45).

(10) Service Water Inlet Temperature Monitoring Instrumentation.

(11) Sampler Flow Rate Monitors.

(12) Boric Acid Makeup Flow System.

(13) Plant Vent Noble Gas Effluent Monitor (R-27).

The amendment also proposes to change the surveillance interval for

the Refueling Water Storage Tank Level to quarterly and to change the

trip setpoint for the Control Rod Protection System. The changes

requested by the licensee are in accordance with Generic Letter 91-04,

``Changes in Technical Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle.''

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) Analog Rod Position Indication:]

The proposed change does not involve a significant hazards

consideration since:

1. A significant increase in the probability or consequences of

an accident previously evaluated will not occur.

It is proposed that the channel calibration frequency for the

analog rod position indication channel be changed from every 18

months (+25%) to every 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that none of the major error contributors are time

dependent and that it can be reasonably expected that the channel

will remain within calibration tolerance over a possible 30 month

operating cycle. In addition, the rod bottom bistable is subject to

monthly testing which would detect any abnormalities in an extended

operating cycle. Due to this monthly test and the acceptable past

test history, it is concluded that the channel will continue to

operate within tolerance over an extended operating cycle and will

not contribute to a significant increase in the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any accident previously evaluated has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval is not expected to affect the

ability of the instrument channel to remain within calibration

tolerance. Furthermore, the rod position indicator is used in normal

operation only as an aid in control rod movement. Normally, very

little control rod movement occurs during normal operation.

Furthermore, it is not relied upon for accident prevention or

accident mitigation. In accordance with existing Technical

Specifications, normal operation can continue even if one channel is

inoperable because alternate means (core instrumentation) exists to

monitor rod position. The frequent monthly test tends to minimize

the effect of a longer operating cycle for the rod position

indication channel as any malfunction induced by time would be

detected. Thus, it is concluded that the possibility of a new or

different kind of accident from any accident previously evaluated

has not been created.

3. A significant reduction in a margin of safety is not

involved.

A statistical analysis of past calibration data has not

identified any time dependent error contributors. Also, past test

data indicates that the channel remains within calibration tolerance

over the existing operating cycle. A longer operating cycle would

increase the risk of drift, however accuracy is not a prime

requirement for the RPI. Therefore, it is concluded that a longer

operating cycle will not result in a significant reduction in a

margin of safety.

[(2) Plant Noble Gas Activity Monitor (R-44):]

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the channel calibration frequency for the

Plant Noble Gas Activity Monitor (R-44) be changed from every 18

months (+25%) to every 24 months (+25%).

The function of R-44 is to respond to high activity levels

during normal operation.

The setpoint for R-44 is established sufficiently above the

expected radioactivity level in the discharge stream to preclude

false actions but sufficiently below the allowed discharge

radioactivity concentration so that discharge in excess of

permissible limits does not occur. Monitor readouts are not used for

quantitative purposes, but are used to respond to relative changes

in radioactivity concentration.

There is limited data to support an unqualified extension of the

surveillance interval. However, the instrument is checked for

operability prior to release. Should the instrument be inoperable

releases may continue provided grab sample analysis is performed.

Since the monitor is subject to daily channel checks, monthly source

checks, and quarterly functional channel tests, abnormal instrument

behavior or inoperability would be detected permitting corrective

actions during the extended surveillance interval.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Operability of the instrument is important rather than ability

to maintain a specific setpoint. Operability of the instrument is

verified prior to a planned discharge and this is independent of an

extended surveillance cycle.

3. There has been no reduction in the margin of safety.

As the Technical Specifications permit pre-planned release even

with an inoperable instrument, the margin of safety is not impacted

by an extended surveillance interval provided that instrument

operability is verified prior to release. This is also required by

the Technical Specifications.

[(3) Low Turbine Auto Stop Oil Pressure Reactor Trip:]

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the channel calibration frequency for the

Low Turbine Auto Stop Oil Pressure system be changed from every 18

months (+25%) to every 24 months (+25%).

No credit is taken for a reactor trip from a low turbine auto

stop oil pressure signal resulting from a turbine trip. Rather, the

safety analysis assumes this reactor trip does not occur during full

load rejection until an overpower delta T condition causes a reactor

trip. In addition, no credit is taken for this system for turbine

missile protection. Therefore, extending the surveillance interval

for this parameter has no impact upon the probability or

consequences of an accident.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

As no credit is taken in the safety analysis for this trip, the

possibility of a new or different kind of accident has not been

created by extending the surveillance interval.

3. There has been no reduction in the margin of safety.

Past test results have not identified any failures. Therefore,

pursuant to Generic Letter 91-04, it is reasonably expected that

this system will continue to function in an acceptable manner over

an extended operating cycle.

[(4) 6.9 kv Undervoltage Relays:]

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the calibration frequency for the 6.9 kv

undervoltage channel be changed from every 18 months (+25%) to every

24 months (+25%).

Quarterly testing of these relays is required by Technical

Specifications. The data from the quarterly tests of the new relays

will be used to assure that drift does not exceed projected values.

The quarterly tests provide a means of maintaining calibration

within specified values, virtually eliminating any impact upon

safety from an extended operating cycle.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Because the quarterly tests assure that relay performance

remains within specified limits, there is no possibility of creating

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

3. There has been no reduction in the margin of safety.

The requirement for a channel functional test each quarter

minimizes any potential impact upon safety due to an extended

operating cycle.

[(5) Boric Acid Tank Level:]

The proposed change does not involve a significant hazards

consideration since:

1. A significant increase in the probability or consequence of

an accident previously evaluated will not occur.

It is proposed that the channel calibration frequency for the

Boric Acid Tank Level instrumentation be changed from every 18

months (+25%) to every 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists between the existing

Technical Specification limit and the licensing basis Safety

Analysis limit to accommodate the channel statistical error

resulting from a 30 month operating cycle. The existing margin

between the Technical Specification limit and the Safety Analysis

limit provides assurance that plant protective actions will occur as

required. It is therefore concluded that changing the surveillance

interval from 18 months (+25%) to 24 months (+25%) will not result

in a significant increase in the probability or consequences of an

accident previously evaluated.

2. The possibility of a new or different kind of accident from

any accident previously evaluated has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds the current margin between the

existing Technical Specification limit and the Safety Analysis

limit. Plant equipment, which will be set at (or more conservatively

than) Technical Specification limits, will provide protective

functions to assure that Safety Analysis limits are not exceeded.

This will prevent the possibility of a new or different kind of

accident from any previously evaluated from occurring.

3. A significant reduction in a margin of safety is not

involved.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds the margin which exists between the current

Technical Specification limit and the licensing basis Safety

Analysis limit. This margin, which is equivalent to the existing

margin, is necessary to assure that protective safety functions will

occur so that Safety Analysis limits are not exceeded.

[(6) Vapor Containment Sump Discharge Flow and Temperature

Channel:]

The proposed change does not involve a significant hazards

consideration since:

1. A significant increase in the probability or consequences of

an accident previously evaluated will not occur.

It is proposed that the calibration frequency for the VC sump

discharge flow and temperature channel be changed from every 18

months (+25%) to every 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists between the existing

Technical Specification and the licensing basis Safety Analysis to

accommodate the channel statistical error resulting from a 30 month

operating cycle. The existing margin between the Technical

Specification and the Safety Analysis provides assurance that plant

protective actions will occur as required. It is therefore concluded

that changing the surveillance interval from 18 months (+25%) to 24

months (+25%) will not result in a significant increase in the

probability or consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any accident previously evaluated has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds the current margin between the

existing Technical Specification and the Safety Analysis. Plant

equipment, which will be set at (or more conservatively than)

Technical Specification limits, will provide protective functions to

assure that Safety Analysis limits are not exceeded. This will

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

any previously evaluated from occurring.

3. A significant reduction in a margin of safety is not

involved.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds the margin which exists between the current

Technical Specification and the licensing basis Safety Analysis.

This margin, which is equivalent to the existing margin, is

necessary to assure that protective safety functions will occur so

that Safety Analysis limits are not exceeded.

[(7) Loss of Power Undervoltage and Degraded Voltage Relays:]

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

The Technical Specifications specify that the Loss of Power

(undervoltage and degraded voltage) relays be calibrated and tested

at a refueling interval; that the undervoltage alarm be calibrated

at a refueling interval, and that the undervoltage (station

blackout) input to Auxiliary Feedwater be calibrated at refueling

intervals. It is proposed that the surveillance frequency be revised

from 18 months (+25%) to 24 months (+25%).

All of the undervoltage and station blackout relays were found

to be within specification at each of the refueling outage

calibration periods.

Since the old relays have been replaced with relays from a

different manufacturer whose drift characteristics are expected to

be superior, extending the surveillance interval by several months

will not significantly increase the probability or consequences of

an accident.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Past test results provide reasonable assurance that the relays

will perform in an acceptable manner for an extended operating

cycle. With the installation of the new relays, whose performance

will surpass the old relays, it is concluded that the plant will

perform within its design basis for an extended operating cycle.

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

from any previously analyzed has not been created.

3. There has been no significant reduction in the margin of

safety.

Since the new relays will surpass the performance of the old

relays, there is reasonable assurance that a significant reduction

in the margin of safety has not resulted from an extended operating

cycle.

[(8) Over-pressurization Protection System (OPS) and Control Rod

Protection System (for use with Low Parasite (LOPAR) fuel) Trip:]

The proposed change does not involve a significant hazards

consideration since:

1. A significant increase in the probability or consequences of

an accident previously evaluated will not occur.

It is proposed that the channel calibration frequency for the

Over-pressurization protection system and the LOPAR trip system be

changed from every 18 months (+25%) to every 24 months (+25%). This

necessitates a change in the LOPAR Technical Specification trip

setpoint from 350 deg.F to 381 deg.F.

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed based upon historical test data.

Based on this analysis, a change to the Technical Specifications is

required. Sufficient margin exists between the Safety Analysis limit

and the proposed Technical Specification limit to accommodate

projected channel uncertainty over a 30 month operating cycle. A

statistical basis exists to assure that protective action will occur

to prevent Safety Analysis limits from being exceeded. Thus, there

will not be a significant increase in the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident

previously evaluated has not been created.

Based upon a statistical analysis of past historical test data

it has been demonstrated that reasonable assurance exists to

conclude that Safety Analysis limits will not be exceeded over a 30

month operating cycle. The proposed Technical Specification limits

provide margin with respect to the Safety Analysis limits and

confidence that appropriate plant protective response will be

provided to prevent the possibility of a new or different kind of

accident from that previously evaluated from being created.

3. A significant reduction in a margin of safety is not

involved.

The proposed changes to the Technical Specification limits are

being made to assure that the previously established margin remains

the same between plant protective function set points and Safety

Analysis limits. This margin is based upon an evaluation of past

historical test data and analytical methods for projecting

instrument channel uncertainty over a 30 month operating cycle. It

is therefore concluded that the existing margin of safety has been

preserved.

[(9) Condenser Evacuation System Activity Monitor (R-45):]

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the channel calibration frequency for the

Condenser Evacuation System Noble Gas Activity Monitor (R-45) be

changed from every 18 months (+25%) to every 24 months (+25%).

Since this radiation monitor is relatively new a degree of

uncertainty is introduced by extending the surveillance interval by

several months. However, the setpoint for automatic diversion is set

some what conservatively. It is established sufficiently high to

avoid spurious actuations and yet sufficiently low so that diversion

and alarm can occur should a step increase in radioactivity level

occur. Under these circumstances considerable departure from the

setpoint can be accommodated and the monitor will still perform its

intended safety function. Continued monitor operability is important

and malfunction would be detected by monthly checks during the

extended operating cycle. Thus, despite the introduction of a new

monitor, the capability of R-45 to tolerate drift in addition to

monthly operator checks, leads to the conclusion that an extended

operating cycle will not result in a significant increase in the

probability or consequences of an accident.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Monthly checks would identify abnormal operating

characteristics, should the instrument fail to perform its intended

function. In the event of tube rupture with a reactor coolant system

radioactivity concentration corresponding to 1% defective fuel, the

resultant site boundary dose would be within 10 CFR [part] 20 limits

should the monitor fail to perform its function (as discussed in

FSAR [Final Safety Analysis Report]). In addition, alternate means

of alarms to indicate a tube rupture event are available. Thus, the

possibility of a new or different kind of accident has not been

created.

3. There has been no reduction in the margin of safety.

Although this monitor is not necessary to mitigate releases

below regulatory limits, it does provide the earliest of a steam

generator tube leak. In this regard, continued instrument operation

is important. Continued instrument operability would be verified by

the monthly checks in an extended operating cycle.

[(10) Service Water Inlet Temperature Monitoring

Instrumentation:]

The proposed change does not involve a significant hazards

consideration since:

1. A significant increase in the probability or consequences of

an accident previously evaluated will not occur.

It is proposed that the channel calibration frequency for the

Service Water Inlet Temperature Monitoring Instrumentation be

changed from every 18 months (+25%) to every 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists between the existing

Technical Specification and the licensing basis Safety Analysis to

accommodate the channel statistical error resulting from a 30 month

operating cycle. The existing margin between the Technical

Specification and the Safety Analysis provides assurance that plant

protective actions will occur as required. It is therefore concluded

that changing the surveillance interval from 18 months (+25%) to 24

months (+25%) will not result in a significant increase in the

probability or consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any accident previously evaluated has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds the current margin between the

existing Technical Specification and the Safety Analysis. Plant

equipment, which will be set at (or more conservatively than)

Technical Specification limits, will provide protective functions to

assure that Safety Analyses are not exceeded. This will prevent the

possibility of a new or different kind of accident from any

previously evaluated from occurring.

3. A significant reduction in a margin of safety is not

involved.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds the existing margin between the current

Technical Specification and the licensing basis Safety Analysis.

This margin, which is equivalent to the existing margin, is

necessary to assure that the protective safety functions occur and

that the Safety Analysis limits are not exceeded.

[(11) Sampler Flow Rate Monitor:]

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the channel calibration frequency for the

Sample Flow Rate Monitors be changed from every 18 months (+25%) to

every 24 months (+25%).

The flow rate monitors are used to estimate the total volume of

air passed through filters. There is no setpoint or safety function

served by these monitors. A high level of radioactivity in the

discharge stream is detected by R-43 and/or R-44.

Insofar as discharge via the unit vent is permissible with the

monitors inoperable, extension of the surveillance interval will

have no impact upon safety.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

As the nuclear safety function is provided by other monitors in

the event of high radioactivity levels in the discharge stream,

extension of the surveillance interval will have no impact upon the

creation of a new or different kind of accident.

3. There has been no reduction in the margin of safety.

These flow monitors are utilized to determine the total air flow

through filters for computational purposes. As adequate measures

(other monitors) exist to prevent the possibility of discharging

radioactivity in excess of applicable limits, there is virtually no

impact upon safety incurred by extending the surveillance interval.

[(12) Boric Acid Makeup Flow System:]

The proposed change does not involve a significant hazards

consideration since:

1. A significant increase in the probability or consequences of

an accident previously evaluated will not occur.

It is proposed that the channel calibration frequency for the

Boric Acid Makeup Flow System be revised from every 18 months (+25%)

to every 24 months (+25%). A statistical analysis of channel

uncertainty for a 30 month operating cycle has been performed. Based

upon this analysis it has been concluded that sufficient margin

exists between the existing Technical Specification limit and the

licensing basis Safety Analysis limit to accommodate the channel

statistical error resulting from a 30 month operating cycle. The

existing margin between the Technical Specification limit and the

Safety Analysis limit provides assurance that plant protective

actions will occur as required. It is therefore concluded that

changing the surveillance interval from 18 months (+25%) to 24

months (+25%) will not result in a significant increase in the

probability or consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any accident previously evaluated has not been created.

The proposed change in operating cycle length due to an

increased surveillance interval will not result in a channel

statistical allowance which exceeds the current margin between the

existing Technical Specification limit and the Safety Analysis

limit. Plant equipment, which will be set at (or more conservatively

than) Technical Specification limits, will provide protective

functions to assure that Safety Analysis limits are not exceeded.

This will prevent the possibility of a new or different kind of

accident from any previously evaluated from occurring.

3. A significant reduction in a margin of safety is not

involved.

The above change in surveillance interval resulting from an

increased operating cycle will not result in a channel statistical

allowance which exceeds the margin which exists between the current

Technical Specification limit and the licensing basis Safety

Analysis limit. This margin, which is equivalent to the existing

margin, is necessary to assure that protective safety functions will

occur so that Safety Analysis limits are not exceeded.

[(13) Plant Vent Noble Gas Effluent Monitor (R-27):]

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the channel calibration frequency for the

Plant Vent Noble Gas Effluent Monitor (R-27) be changed from every

18 months (+25%) to every 24 months (+25%).

R-27 is a high range noble gas monitor intended for use after an

accident to provide information about the magnitude of radioactive

releases. It serves no purpose during normal operation. It provides

no function to prevent or mitigate an accident but does provide a

role in assessing the consequences of an accident. As the monitor is

a high range monitor, an estimate of the magnitude of release rather

than accuracy is important. Accordingly, continued operability of

the instrument during an extended operating cycle is more important

than the device exhibiting minimal drift characteristics.

Malfunction of the instrument would be detected by the shift checks

and functional tests performed during the extended operating cycle.

2. The possibility of a new or different kind of accident from

any previously analyzed has not been created.

Since the monitor provides no preventive or mitigating action in

the event of an accident, no new or different type of accident has

been created by extending the operating cycle. In terms of post

accident assessment capability, alternate means exist to assess

offsite releases in the event of failure of this instrument.

3. There has been no reduction in the margin of safety.

Since the instrument provides no safety function and alternate

means exist for post accident assessment purposes, there will be no

impact on safety due to an extended period between calibrations.

[(14) Refueling Water Storage Tank Level:]

The proposed change does not involve a significant hazards

consideration since:

1. A significant increase in the probability or consequences of

an accident previously evaluated will not occur.

It is proposed that the channel calibration frequency for the

RWST instrumentation be changed from every 18 months (+25%) to

quarterly (once every 3 months).

A statistical analysis of channel uncertainty for a 3 month

surveillance has been performed. Based upon this analysis it has

been concluded that sufficient margin exists between the existing

Technical Specification limit and the licensing basis Safety

Analysis limit to accommodate the channel statistical error

resulting from a 3 month quarterly surveillance. The existing margin

between the Technical Specification limit and the Safety Analysis

limit provides assurance that plant protective actions will occur as

required. It is therefore concluded that changing the surveillance

interval from 18 months (+25%) to quarterly will not result in a

significant increase in the probability or consequences of an

accident previously evaluated.

2. The possibility of a new or different kind of accident from

any accident previously evaluated has not been created.

The proposed change in surveillance interval will result in a

channel statistical allowance which provides the necessary margin

between the existing Technical Specification limit and the Safety

Analysis limit. Plant equipment, which will be set at (or more

conservatively than) Technical Specification limits, will provide

protective functions to assure that Safety Analysis limits are not

exceeded. This will prevent the possibility of a new or different

kind of accident from any previously evaluated from occurring.

3. A significant reduction in a margin of safety is not

involved.

The above change in surveillance interval will result in a

channel statistical allowance which is necessary between the current

Technical Specification limit and the licensing basis Safety

Analysis limit. This margin is necessary to assure that protective

safety functions will occur so that Safety Analysis limits are not

exceeded.

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: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for Licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Project Director: Robert A. Capra.

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

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

Date of Amendment Request: March 24, 1994.

Description of Amendment Request: The changes are in support of the

forthcoming Cycle 7 for Catawba, Unit 2. The proposed Technical

Specification (TS) changes reflect:

(1) An increase from 2000 parts per million (ppm) to 2175 ppm in

the required spent fuel storage pool minimum boron concentration during

Modes 1-3 operation,

(2) An increase from 2000 ppm to 2175 ppm in the required reactor

coolant system (RCS) and refueling canal minimum boron concentration

during Mode 6 operation,

(3) The inclusion of two reload related topical reports into TS

6.9.1.9, and

(4) The revision of an administrative nature to correct errors in

nomenclature and to remove obsolete footnotes.

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:

Increase in Boron Concentration Limit for the Spent Fuel Storage Pool

(Standby Makeup Pump Water Supply)

The required spent fuel storage pool minimum boron concentration

was increased from 2000 ppm to 2175 ppm during Modes 1-3.

The proposed revision is conservative, and is required only to

maintain consistency between the boron concentration of the spent

fuel storage pool and the boron concentration of the RWST [refueling

water storage tank] during Modes 1-3 operation. Therefore, there

will be no adverse impact upon the probability or consequences of

any previously analyzed accident.

Likewise, the proposed change will not create the possibility of

a new or different kind of accident, since no new failure modes are

identified.

Finally, no negative impact upon any safety margin is created

since the proposed change is conservative.

Increase in Boron Concentration Limits for the RCS and Refueling Canal

in Mode 6

The increase in the required RCS and refueling canal minimum

boron concentration was added only to maintain consistency between

the boron concentration of the RCS and refueling canal and the RWST

in Mode 6.

The change in boron concentration limits for the RCS and

refueling canal will not increase the probability of an accident

since no accident initiators are involved with this change. Since

the change is conservative, the consequences of an accident

previously evaluated will not be increased. The increase in the

boron concentration limit for the RCS and refueling canal in Mode 6

adds further margin to the initial conditions assumed for the boron

dilution accident in the safety analysis. Therefore, the

consequences of the boron dilution accident previously evaluated

will not be increased.

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

previously evaluated will not be created since this change is

bounded by previously evaluated accidents and does not introduce any

new failure modes.

This change does not involve a significant reduction in the

margin of safety since the analyses performed demonstrate that the

limits imposed meet all accident analysis and design basis

requirements.

Addition of Two Reload Related Topical Reports

This change is administrative in nature and adds two previously

approved topical reports to the list of methodologies used to

determine core operating limits. The change will have no impact upon

either the probability or consequences of a previously analyzed

accident. The methodologies described in the topical reports have

been previously reviewed and approved by the NRC. Also, no new

accident possibilities are created, since this is an administrative

change. Finally, no impact upon any safety margin is created, since

the change is administrative in nature and the described topical

reports have received full NRC approval.

Correction of Errors in Nomenclature and Removal of Obsolete Footnotes

These changes are also administrative in nature and are intended

to correct miscellaneous errors and obsolete references. As such,

the changes will have no impact upon either the probability or

consequences of any previously analyzed accidents, will not create

the possibility of any new accident scenarios, and will not impact

any safety margins.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730.

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

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: David B. Matthews.

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

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

Date of Amendment Request: March 23, 1994.

Description of Amendment Request: The proposed amendments would

revise Technical Specification (TS) 6.9.2, ``Core Operating Limits

Report,'' to include a reference to a Duke Power Company (DPC) Topical

Report describing an analytical method for determining the core

operating limits.

Specifically, the amendments would add: ``(4) DPC-NE-1004A, Nuclear

Design Methodology Using CASMO-3/SIMULATE-3P, November 1992,'' to TS

6.9.2.

Basis for Proposed No Significant Hazards Consideration

Determination: The NRC staff reviewed Topical Report DPC-NE-1004A and

concluded in a Safety Evaluation Report dated November 23, 1992, that

the described nuclear design methodology is acceptable for performing

reload analyses for the DPC B&W 177-assembly cores in the Oconee units.

The addition of this approved nuclear design methodology to those

referenced in TS 6.9.2 provides an alternative method for determining

core operating limits such that all applicable limits (e.g., fuel

thermal mechanical limits, core thermal hydraulic limits, ECCS limits,

nuclear limits such as shutdown margin, and transient and accident

analysis limits) of the safety analysis are met. Therefore, the

proposed change to the TS (1) does not involve a significant increase

in the probability or consequences of an accident previously evaluated,

(2) does not create the possibility of a new or different kind of

accident than previously evaluated, and (3) does not involve a

significant reduction in the margin of safety.

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:

Duke Power Company (Duke) has made the determination that this

amendment request involves a No Significant Hazards Consideration by

applying the standards established in 10 CFR 50.92. This ensures 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:

Each accident analysis addressed within the Oconee Final Safety

Analysis Report (FSAR) has been examined with respect to this

amendment request. The Technical Specifications will continue to

require operation within the bounds of the cycle-specific parameter

limits. The cycle-specific parameter limits will be calculated using

NRC approved methodology. The proposed amendment is simply an

administrative change to update the list of NRC approved methods in

Technical Specification 6.9.2. Therefore, the probability of any

Design Basis Accident (DBA) is not affected by this change, nor are

the consequences of a DBA affected by this change. This is because

the addition of an NRC approved reference to Technical Specification

6.9.2 is not considered to be an initiator or contributor to any

accident analysis addressed in the Oconee FSAR.

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

accident from any kind previously evaluated:

Operation of ONS [Oconee Nuclear Station] in accordance with

these Technical Specifications will not create any failure modes not

bounded by previously evaluated accidents. Consequently, this change

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

accident from any kind of accident previously evaluated.

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

The Technical Specifications will continue to require operation

within the bounds of the cycle-specific parameter limits. Duke will

continue to calculate the cycle-specific parameter limits using NRC

approved methodology. In addition, each future reload will require a

10 CFR 50.59 safety review to ensure that operation of the unit

within the cycle-specific limits will not involve a reduction in a

margin of safety. Therefore, no margins of safety are affected by

the addition of an NRC approved methodology to Technical

Specification 6.9.2.

Based on the staff's analysis and its review of the licensee's

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691.

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

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

NRC Project Director: David B. Matthews.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and

50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,

Georgia

Date of Amendment Request: March 18, 1994.

Description of Amendment Request: The proposed amendments would

revise Technical Specification (TS) 3/4.3.3.6, Accident Monitoring

Instrumentation, TS 3/4.6.4.1, Hydrogen Monitors, and their associated

bases to incorporate the technical substance of Specification 3.3.3

from NUREG-1431, Revision O (Standard Technical Specifications) for the

Westinghouse Owners Group.

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 proposed changes affect instrumentation that would be used to

assess the condition of the plant during and following an accident.

As such, the changes can have no effect on the probability of any

accident previously evaluated since this instrumentation has no

bearing on initiating events. The proposed changes will continue to

ensure the capability to monitor plant conditions during and

following an accident by requiring redundancy or diversity and

timely corrective action in the event of inoperable instrumentation.

Therefore, the proposed changes will not significantly increase the

consequences of any accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated. The proposed changes affect the operability and action

requirements for the post accident monitoring instrumentation

system. Accordingly, the proposed changes do not involve any change

to the configuration or method of operation of any plant equipment,

and no new failure modes have been defined for any plant system or

component nor has any new limiting failure been identified as a

result of the proposed changes. Therefore, 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. The intent of the existing TS requirements is

to ensure the capability to monitor the plant condition during and

following an accident so that the operators will have the

information necessary to monitor and evaluate the course of the

event and take any necessary action. Under the proposed changes this

capability will be maintained by ensuring redundancy or diversity

and by requiring timely corrective action in the event of inoperable

instrumentation. In addition, the proposed changes would avoid

unnecessary plant shutdowns by specifying an appropriate level of

action in response to inoperable instrumentation. 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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

Attorney for Licensee: Mr. Arthur H. Domby, Troutman Sanders,

NationsBank Plaza, suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia 30308.

NRC Project Director: David B. Matthews.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of Amendment Request: April 6, 1994.

Description of Amendment Request: The proposed amendment changes

the Technical Specifications to eliminate the main steam line radiation

monitor(s) (MSLRMs) reactor scram and isolation functions of the MSLRMs

currently contained in Tables 3.1.-1 and 4.1-1 of the Technical

Specifications and the associated Bases statements. This action follows

the recommendations of the BWR Owners Group (BWROG) in their Safety

Evaluation, NEDO-31400A, previously approved by the NRC Staff on May

15, 1991 by letter to the BWROG. Following is a brief description of

the proposed changes:

Tech. Spec. 3.1, ``Protective Instrumentation'' Bases is revised to

delete reference to the paragraph describing the Main Steam Line (MSL)

radiation monitoring functions for indication of excessive fuel failure

and initiation of a reactor scram and MSL isolation.

Tech. Spec. Table 3.1.1., ``Protective Instrumentation Requirements

- A. Reactor Scram Functions,'' is revised to delete line Item No. 7 -

``High Radiation in Main Steam Line Tunnel.''

Tech. Spec. Table 3.1.1., ``Protective Instrumentation Requirements

- B. Reactor Isolation Functions,'' is revised to delete line Item No.

6 - ``High Radiation in Main Steam Line Tunnel.''

Tech. Spec. Table 3.1.1., ``Protective Instrumentation Requirements

- L. Condenser Vacuum Pump Isolation Function,'' is revised to delete

line Item No. 1 - ``High Radiation in Main Steam Line Tunnel.''

Tech. Spec. Table 4.1.1., ``Minimum Check, Calibration and Test

Frequency For Protective Instrumentation,'' is revised to delete

Instrumentation Channel No. 13 - ``High Radiation in Main Steam Line.''

Basis for Proposed No Significant Hazards Consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated.

The objective of the MSLRMs is to provide early indication of

gross fuel failure. The monitors provide an alarm function, and

signals that lead to a scram function and [main steam isolation

valve] MSIV isolation functions. The basis for the MSIV isolation on

an MSL high radiation signal is to reduce the quantity of fission

products transported from the reactor vessel to the condenser in the

event of gross fuel failure. No [design basis accident] DBA takes

credit for a reactor scram resulting from an MSL high radiation

signal.

The proposed change removes all trip functions of the MSLRMs.

The only modification attendant to this change is the removal of

contacts derived from the MSLRM logic to the reactor scram, reactor

isolation and offgas system isolation initiation logic. This change

does not affect the operation of any equipment having the potential

to cause a [control rod drop accident] CRDA. Therefore, the

probability of a CRDA is not increased or in any way affected by the

proposed change.

However, the CRDA analysis does take credit for MSIV isolation.

As discussed above, assuming no MSIV isolation in the event of a

CRDA, the offsite radiation doses will remain a small fraction of

the 10 CFR part 100 Reactor Site Criteria.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The function of an MSLRM trip is to detect abnormal fission

product release and isolate the steam lines, thereby stopping the

transport of fission products from the reactor to the main

condenser. No credit is taken for the reactor scram function due to

the action of these monitors on high radiation in the MSLs in any

design basis accident. Removing the MSLRMs MSL isolation trip and

its subsequent reactor scram will not affect the operation of other

equipment or systems necessary for the prevention or mitigation of

accidents.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

Eliminating the MSLRM trip functions as analyzed in NEDO-31400A

will result in a potential increase in the margin of safety because

of:

a. Improvement in the availability of the main condenser for

decay heat removal; and,

b. Elimination of inadvertent reactor scrams and challenges to

safety systems.

Therefore, operation of the facility in accordance with the

proposed changes will not result in a reduction of safety margin.

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 No. 50-316, Donald C. Cook

Nuclear Plant, Unit No. 2, Berrien County, Michigan

Date of Amendment Request: March 9, 1994

Description of Amendment Request: The proposed amendment would

modify the Technical Specification to allow a one time exemption from

certain Appendix J testing. This exemption would extend the interval

for Type B and C testing until the Unit 2 refueling outage currently

scheduled for August 1994.

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:

As stated in 10 CFR 50.92(c), 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, or (2) the change does not create

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

accident previously evaluated, or (3) the change does not involve a

significant reduction in a margin of safety.

Criterion 1

The limiting conditions for operation involving containment

integrity are not altered by this proposed change. The surveillance

requirement concerning the Type B and C leak rate test is slightly

relaxed by the proposed change. The function of the components

affected by this surveillance are to ensure containment integrity.

Delaying the surveillance approximately two months would not change

the probability of an accident. Our significant improvement in Type

B and C leak rate test results, low anticipated leak rate for the

next surveillance, aggressive corrective actions taken, and

excellent ILRT [integrated leak rate test] results indicate there is

no reason to believe that delaying the Type B and C leak rate tests

approximately two months will cause serious deterioration to these

components. Furthermore, similar requests by utilities to extend the

surveillance beyond two years 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 for

containment integrity are proposed as part of this amendment

request. The proposed change does not involve any physical changes

to the plant or any changes to plant operations. Thus, the proposed

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

accident from any accident previously evaluated.

Criterion 3

The intent of the Type B and C leak rate surveillance is to

ensure that containment integrity does not significantly

deteriorate. This is established by measuring a total leak rate of

less than 0.60 La. Our significant improvement in Type B and C

leak rate tests results, aggressive corrective actions taken, and

excellent ILRT results indicate there is no reason to believe that

delaying the Type B and C leak rate tests approximately two months

will cause serious deterioration to these components. The ``As

Found'' trend of the leak rates over the past three surveillances

indicate that the leak rate for the next surveillance will be below

the Appendix J leak rate acceptance criteria. Therefore, it is

concluded that the proposed amendment does not involve a reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: 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: Ledyard B. Marsh.

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

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

Date of Amendment Request: March 23, 1994.

Description of Amendment Request: The licensee proposed to modify

Technical Specification Table 3.7-6, Area Temperature Monitoring, by

creating two zones for the main steam valve building (MSVB) and

increasing the maximum normal excursion (MNE) temperature limit for

this area from 120 deg.F to 140 deg.F.

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 change does not involve an SHC [significant hazards

consideration] because the change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The increase of the MNE temperature from 120 deg.F to 140

deg.F for the main steam valve building has been evaluated. The

equipment in the building has been shown to be qualified for

continuous operation at 140oF. The effect of this temperature change

has decreased slightly the qualified life of the components in the

building. For those components with a qualified life of less than 40

years, they will be replaced as a scheduled maintenance item.

An engineering review of the MSLB profile for this building was

conducted and it was concluded that those components required to

operate post accident, will continue to perform their safety

function. Therefore, since the equipment will continue to operate as

designed both during normal conditions and subsequent to a MSLB, the

probability or consequences of an accident previously evaluated is

not increased.

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

from any previously analyzed.

The effect of increasing the MNE temperature to 140 deg.F has

been evaluated and judged acceptable. The possible failure of the

equipment in this building due to the increase in temperature is no

more likely than it was before, since the equipment has been shown

to be qualified to 140 deg.F. Failure of any equipment in this

building at the new temperature will not create any new accidents or

consequences that were not considered previously.

Finally, since there are no changes in the way the plant is

operated, there is no possibility of an accident of a new or

different type than previously evaluated due to the proposed change.

3. Involve a significant reduction in margin of safety.

The proposed change increases the MNE temperature within the

MSVB. The equipment in the building has been reviewed to ensure

operability. There is a slight decrease in the qualified life, but

this was anticipated and scheduled previously and any such

replacement of equipment will continue as a maintenance item. A

review of the MSLB profile was performed for this area and it was

shown that the required equipment will continue to operate as

required.

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.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of Amendment Request: March 15, 1994.

Description of Amendment Request: The proposed amendment would

include the use of integral fuel burnable absorbers as a method of

controlling core excess reactivity and maintaining the core power

distribution within acceptable peaking limitations.

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 staff's review is

presented below:

1. The proposed amendment would not involve a significant increase

in the probability or consequences of an accident previously evaluated.

Any fuel containing integral burnable absorbers will be analyzed

using NRC approved methods and acceptance criteria prior to being

loaded into Maine Yankee's reactor vessel core. Verification of

adequate shutdown margin is performed during low power physics testing

after each refueling. In addition, core physics monitoring is required

during power operation by Technical Specifications sections 3.10, ``CEA

Group, Power Distribution, Moderator Temperature Coefficient Limits and

Coolant Conditions,'' and 3.15 ``Reactivity Anomalies.'' Such testing

and monitoring ensures adequate margin exists to accommodate the

anticipated transients and accidents postulated in Maine Yankee's Final

Safety Analysis Report.

The licensee therefore concludes that implementation of the

proposed change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. The proposed amendment would not create the possibility of a new

or different kind of accident from any accident previously evaluated.

A determination of compliance with approved acceptance criteria is

made for every Maine Yankee fuel reload prior to loading fuel. The use

of approved methodologies and acceptance criteria ensure that new or

different accidents will not be created by the use of integral fuel

burnable absorbers.

The licensee therefore concludes that implementation of the

proposed change will not create any or new or different kind of

accident from any accident previously evaluated.

3. The proposed amendment would not involve a significant reduction

in a margin of safety.

The safety evaluation performed for each core reload ensures that

the core design meets appropriate acceptance criteria. Because these

criteria remain unchanged as approved by the NRC, the margin of safety

remains the same.

The licensee therefore concludes that implementation of the

proposed change would not involve a significant reduction in a margin

of safety.

Based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room Location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, Maine 04578.

Attorney for Licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 83 Edison Drive, Augusta, Maine 04336.

NRC Project Director: Walter R. Butler.

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

Generating Plant, Wright County, Minnesota

Date of Amendment Request: January 26, 1994.

Description of Amendment Request: The proposed amendment would

revise the Technical Specifications and associated Bases to reflect the

fact that the main steam isolation valves can now be tested at a

pressure of greater than or equal to Pa (42 psig) thereby

eliminating the need for the previously granted exemption to certain

Appendix J testing requirements. The exemption would no longer be

necessary because of improvements in testing technology.

Basis for Proposed No Significant Hazards Consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

a. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment is limited to changes to the surveillance

testing requirements (test pressure and allowable leakage criteria)

applicable to the main steam line isolation valves. The proposed

criteria are equivalent to the current criteria with respect to

monitoring main steam isolation valve performance to ensure that

leakage past the valves would be within acceptable limits under

accident conditions. This surveillance test is performed while the

plant is in a cold shutdown condition at a time when the main steam

isolation valves are not required to be operable. Performance of the

test itself is not an input or consideration in any accident

previously evaluated, thus the proposed change will not increase the

probability of any such accident occurring.

The proposed amendment will not adversely affect the function,

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

capability of the valves to perform as required during an accident.

There will be no increase in post accident off-site or on-site

radiation dose, since the adjusted leakage limit is consistent with

inputs previously established for the dose analyses. The proposed

amendment is consistent with regulatory requirements (10 CFR Part

50, Appendix J) and guidance (TER-C5257-30) that has been previously

reviewed by the NRC and found to be acceptable. Therefore, the

amendment will not increase the consequences of any accident

previously evaluated.

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

new or different kind of accident from any accident previously

analyzed.

The proposed amendment does not involve any modification to

plant equipment or operating procedures, nor will it introduce any

new main steam isolation valve failure modes that have not been

previously considered. The proposed amendment is limited to a change

in the surveillance test pressure & acceptance criteria used to leak

test the valves. This test is performed while the plant is in a cold

shutdown condition at a time when the valves are not required to be

operable. We therefore conclude the proposed changes will not create

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

accident previously analyzed.

c. The proposed amendment will not involve a significant

reduction in the margin of safety.

The proposed amendment will result in the main steam isolation

valves being subjected to the maximum pressure (Pa, 42 psig)

calculated to occur under worst case accident conditions, and will

therefore provide a more realistic and challenging test of valve

performance under those conditions. The leakage rate criteria for

the test has been adjusted upward to be commensurate with the higher

test pressure, but this does not represent any increase in actual

leakage under accident conditions. On-site and off-site dose

analyses will not be affected. The proposed amendment does not

involve any change in operability requirements or limiting

conditions for operation beyond the replacement of the old test

pressure & acceptance criteria with equivalent criteria consistent

with 10 CFR Part 50, Appendix J, NUREG-1433, and TER-C5257-30. Based

on these considerations, we conclude the proposed amendment will not

involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

NRC Project Director: Ledyard B. Marsh.

Philadelphia Electric Company, Public Service Electric and Gas Company,

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

Docket No. 50-277, Peach Bottom Atomic Power Station, Unit No. 2, York

County, Pennsylvania

Date of Application for Amendment: April 6, 1994.

Description of Amendment Request: The amendment would reflect the

incorporation of the end-of-cycle Minimum Critical Power Ratio

Recirculation Pump Trip (MCPR-RPT) system and the replacement of the

Reactor Recirculation System (RRS) Motor Generator (M-G) Sets with

solid state adjustable speed drives (ASDs).

Basis for Proposed No Significant Hazards Consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

The addition of the end-of-cycle MCPR-RPT System, which utilizes

ASDs, will not have a significant increase in the probability or

consequences of an accident previously evaluated.

The end-of-cycle MCPR-RPT System has been designed to

appropriate standards and specifications to ensure that the ability

of the plant to mitigate the effects of accidents is maintained.

Additionally, the MCPR-RPT System has been analyzed such that no new

accident initiators will be created such that the probability of an

accident previously evaluated will not increase.

No new challenges to the reactor coolant pressure boundary will

result from the incorporation of the end-of-cycle MCPR-RPT System

which could result in an increase in the consequences of an

accident. All engineered safety features will function as described

in the PBAPS UFSAR [Peach Bottom Atomic Power Station Updated Final

Safety Analysis Report] in order to mitigate the consequences of

accidents previously evaluated in the PBAPS UFSAR. Additionally, all

fission product barriers and safety margins will be maintained.

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

or different kind of accident from any previously evaluated.

The end-of-cycle MCPR-RPT System, which utilizes ASDs, has been

designed to appropriate standards and specification to ensure that

no new sequence of events or failure modes will occur such that a

transient event will escalate into a new or different type of

accident.

The software used in the digital system of the ASDs is not

subject to the verification and validation requirements discussed in

the NRC memorandum dated July 1, 1991, from A. C. Thadoni [sic]

[Thadani] (NRC) to S. A. Varga (NRC) and B. A. Bolger [sic] [Boger]

(NRC), because this equipment is neither safety-related nor

important to safety. There is no software used in the trip circuit

of the end-of-cycle MCPR-RPT System, except for the ASDs.

Additionally, the design of the modification will assure that the

new equipment EM emissions will not cause inadvertent operation of

existing plant equipment and that harmonic filters have been

incorporated to minimize electrical noise on the 13kV input power

buses.

(3) The proposed change does not result in a significant

reduction in the margin of safety.

The incorporation of the end-of-cycle MCPR-RPT System, which

utilizes ASDs, will not result in a reduction in the margin of

safety. All safety margins will be maintained.

The end-of-cycle MCPR-RPT System will aid in protecting the

integrity of the fuel barrier by tripping the recirculation pumps

early in the pressurization phase of the load rejection with no

bypass event, the turbine trip with no bypass event, and the

feedwater controller failure--maximum demand event. The early

tripping of the recirculation pumps will introduce negative void

reactivity thus reducing reactor power and maintaining safety

margins. The end-of-cycle MCPR-RPT System will ensure CPR safety

margins which protect fuel barrier integrity.

General Electric has performed a qualitative assessment of

transients that would be impacted as a result of replacing the M-G

Sets with ASDs. General Electric concluded that the faster coastdown

of the recirculation pumps during a Loss of Coolant Accident (LOCA)

due to the removal of the M-G Set inertia may slightly increase the

peak clad temperature during this event. This increase is expected

to be less than 50 deg.F. The small increase will not exceed the

2200 deg.F peak cladding temperature regulatory limit. No design or

safety limit will be exceeded.

The replacement of the M-G Sets with the ASDs will not impact

the recirculation flow controller failure--increase flow transient.

The UFSAR analysis assumes a 25%/sec rate of increase. The ASD

control system will include rate limiters that prevent a pump speed

increase greater than 25%/sec in the event of a failure. Thus, the

consequences of this transient remain bounded and safety margins

will be maintained.

The ASDs will also allow a ``soft start'' of the recirculation

pumps with the recirculation discharge valves closed prior to pump

start and a gradual increase in pump speed. This results in a

gradual change in core flow. Thus, the response to a startup of an

idle recirculation pump remains bounded by the transient analysis

and safety margins will be maintained in the transient analyses.

Changes to the fire protection equipment will still maintain the

capability to shutdown the plant in the event of a fire.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Government Publications

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

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

Pennsylvania 17105.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: Charles L. Miller.

Philadelphia Electric Company, Public Service Electric and Gas Company,

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

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

Units Nos. 2 and 3, York County, Pennsylvania

Date of Application for Amendments: March 28, 1994.

Description of Amendment Request: The proposed Technical

Specifications (TS) changes relocate the TS fire protection

requirements to the Updated Final Safety Analysis Report.

Basis for Proposed No Significant Hazards Consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes are administrative in nature and are

consistent with the guidance provided in NRC GL's [Generic Letters]

86-10 and 88-12. They do not affect the initial conditions or

precursors assumed in the Updated Final Safety Analysis Report

Section 14. These changes do not decrease the effectiveness of

equipment relied upon to mitigate the previously evaluated

accidents.

Therefore, there is no increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any previously evaluated.

The proposed changes do not make any physical changes to the

plant or changes to operating procedures. Therefore, implementation

of the proposed changes will not affect the design function or

configuration of any component or introduce any new operating

scenarios or failure modes or accident initiation.

Therefore, the proposed changes do not create the possibility of

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

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes are administrative in nature and are

consistent with the guidance provided in NRC GL's 86-10 and 88-12.

The proposed changes do not adversely affect the assumptions or

sequence of events used in any accident analysis.

Therefore, the proposed changes do not involve a reduction in

any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Government Publications

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

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

Pennsylvania 17105.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: Charles L. Miller.

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

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

Jersey

Date of Amendment Request: January 21, 1994.

Description of Amendment Request: These amendments would revise

Technical Specifications 3.8.2.3 for both Salem Unit 1 and Salem Unit 2

to include the battery acceptance criteria, corresponding allowed

outage times and additional surveillance requirements recommended in

NUREG-1431, Standard Technical Specifications--Westinghouse Plants.

TS 3.8.2.4 ``125 Volt D.C. Distribution--Shutdown'' would also be

indirectly affected by these changes because it refers to the

surveillance requirements of TS 4.8.2.3.2 to demonstrate the battery

and chargers Operable.

In addition, Salem Unit 1 TS 3.8.2.3 Limiting Condition for

Operation (LCO) would be revised to define the specific battery charger

required for each train. Salem Unit 1 TS 3.8.2.3 Action Statement would

also be revised to restrict the use of the backup battery charger to a

period not to exceed 7 days.

Additionally, the Unit 1 action statement for an inoperable 125

volt DC bus would be modified to add the requirement that the bus also

be energized.

Basis for Proposed No Significant Hazards Consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes do not alter plant configuration or

operation. The proposed changes do not invalidate any of the

parameters assumed in the plants UFSAR Design Basis Accident or

Transient Analyses. The proposed changes provide additional guidance

to be used to ensure operability of the safety related batteries.

New surveillance requirements and specific battery cell parameters

offer improved monitoring of the battery status. The new guidance

and surveillance requirements are consistent with the

recommendations of NUREG-1431, Standard Technical Specifications--

Westinghouse Plants, and current industry recommendations.

The changes to the Unit 1 LCO and corresponding Action Statement

restrict the use of the backup battery charger, thereby limiting the

amount of time that one AC Vital bus is allowed to power the

chargers of more than one DC train. This change brings the TS for

both Units into agreement and results in a more conservative Unit 1

TS.

Therefore, the probability or consequences of an accident

previously evaluated are not increased by the proposed change.

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

from any accident previously evaluated.

The proposed changes do not introduce any design or physical

configuration changes to the facility or change the method by which

any safety-related system performs its function. The proposed

changes are consistent with the recommendations of NUREG-1431,

Standard Technical Specifications--Westinghouse Plants. Therefore,

the proposed changes will not increase the possibility of a new or

different kind of accident from any accident previously identified.

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

safety.

The proposed changes do not alter the manner in which safety

limits or limiting safety system setpoints are determined. The new

cell parameter table and additional surveillance requirements

provide improved means to monitor and evaluate overall battery

performance. Therefore, the proposed changes do not involve a

significant reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079.

Attorney for Licensee: Mark J. Wetterhahn, Esquire, Winston and

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

NRC Project Director: Charles L. Miller.

South Carolina Electric & Gas Company, South Carolina Public Service

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

No. 1, Fairfield County, South Carolina

Date of Amendment Request: October 29, 1993.

Description of Amendment Request: The licensee is preparing to

replace the currently installed steam generators with new model Delta

75 steam generators (SGs). The new steam generators will be larger than

those currently installed. The physical changes to the plant and the

accident reanalyses needed to support those changes will necessitate

increasing the maximum tested charging/safety injection pump flow rate

from 680 gallons per minute to 700 gallons per minute.

Basis for Proposed No Significant Hazards Consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Operation of VCSNS [Virgil C. Summer Nuclear Station] in

accordance with the proposed license amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Implementation of the [Delta] 75 SGs and revised operating

conditions do not contribute to the initiation of any accident

evaluated in the FSAR [Final Safety Analysis Report]. Supporting

factors are as follows:

--The [Delta] 75 SG is designed in accordance with ASME [American

Society of Mechanical Engineers] Code Section III, 1986 edition

[sic] and other applicable federal, state, and local laws, codes and

regulations and meets the original interfaces for the Model D3 SGs

with exception that provisions for a larger blowdown nozzle have

been made and the feedwater inlet nozzle is located in the upper

shell.

--All NSSS [nuclear steam supply system] components (i.e., reactor

vessel, RC Pumps, pressurizer, CRDM's [control rod drive

mechanisms], [Delta] 75 SGs, and RCS piping) are compatible with the

revised operating conditions. Their structural integrity is

maintained during all proposed plant conditions through compliance

with the ASME code.

--Fluid and auxiliary systems which are important to safety,

including the CHG/SI [charging and safety injection] system with

maximum pump flows up to 700 gpm, are not adversely impacted and

will continue to perform their design function.

--Overall plant performance and operation are not significantly

altered by the proposed changes.

Therefore, since the reactor coolant pressure boundary integrity

and system functions are not adversely impacted, the probability of

occurrence of an accident evaluated in the VCSNS FSAR will be no

greater than the original design basis of the plant.

An extensive analysis has been performed to evaluate the

consequences of the following accident types currently evaluated in

the VCSNS FSAR:

--Non-LOCA [non-loss-of-coolant accident]

--Large Break and Small Break LOCA

--Steam Generator Tube Rupture

With the [Delta] 75 SGs and revised operating conditions, the

calculated results (i.e., DNBR [departure from nucleate boiling

ratio], Primary and Secondary System Pressure, Peak Clad

Temperature, Metal Water Reaction, Challenge to Long Term Cooling,

Environmental Conditions Inside and Outside Containment, etc.) for

the accidents are similar to those currently reported in the VCSNS

FSAR. Select results (i.e., Containment Pressure during a Steam Line

Break, Minimum DNBR for Rod Withdrawal from Subcritical, etc.) are

slightly more limiting than those reported in the current FSAR due

to the use of the assumed operating conditions with the new [Delta]

75 SGs, and in some cases, use of an uprated core power of 2900 MWt.

However, in all cases, the calculated results do not challenge the

integrity of the primary/secondary/containment pressure boundary and

remain within the regulatory acceptance criteria applied to VCSNS's

current licensing basis. The assumptions utilized in the

radiological evaluations, described in Section 3.7, are thus

appropriate and are judged to provide a conservative estimate of the

radiological consequences during accident conditions. Given that

calculated radiological consequences are not significantly higher

than current FSAR results and remain well within 10CFR100 limits, it

is concluded that the consequences of an accident previously

evaluated in the FSAR are not increased.

(2) The proposed license amendment does not create the

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

previously evaluated.

The [Delta] 75 SGs, revised operating conditions, and higher

allowable CHG/SI pump flows will not introduce any new accident

initiator mechanisms. Structural integrity of the RCS is maintained

during all plant conditions through compliance with the ASME code.

No new failure modes or limiting single failures have been

identified. Design requirements of auxiliary systems are met with

the RSGs [Replacement Steam Generators]. Since the safety and design

requirements continue to be met and the integrity of the reactor

coolant system pressure boundary is not challenged, no new accident

scenarios have been created. Therefore, the types of accidents

defined in the FSAR continue to represent the credible spectrum of

events to be analyzed which determine safe plant operation.

(3) The proposed license amendment does not involve a

significant reduction in a margin of safety.

Although the [Delta] 75 SGs, revised operating conditions, and

higher allowable CHG/SI pump flows will require changes to the VCSNS

Technical Specifications, it will not invalidate the LOCA, non-LOCA,

or SGTR [steam generator tube rupture] conclusions presented in the

FSAR accident analyses. For all the FSAR non-LOCA transients, the

DNB design basis, primary and secondary pressure limits, and dose

limits continue to be met. The LOCA peak cladding temperatures

remain below the limits specified in 10 CFR 50.46. The calculated

doses resulting from a SGTR event will continue to remain within a

small fraction of the 10 CFR 100 permissible releases. Environmental

conditions associated with High Energy Line Break (HELB) both inside

and outside containment have been evaluated.

The containment design pressure will not be violated as a result

of the HELB. Equipment qualification will be updated, as necessary,

to reflect the revised conditions resulting from HELB. The margin of

safety with respect to primary pressure boundary is provided, in

part, by the safety factors included in the ASME Code. Since the

components remain in compliance with the codes and standards in

effect when VCSNS was originally licensed (with the exception of the

[Delta] 75 RSGs which use the 1986 ASME Code Section III Edition),

the margin of safety is not reduced. Thus, there is no reduction in

the margin of safety as defined in the bases of the VCSNS Technical

Specifications.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Fairfield County Library,

Garden and Washington Streets, Winnsboro, South Carolina 29180.

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

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

NRC Project Director: William H. Bateman.

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power

Plant, Unit No. 1, Lake County, Ohio

Date of Amendment Request: March 12, 1993.

Description of Amendment Request: The proposed amendment would

revise Technical Specification Table 3.3.7.1-1, to clarify the actions

to be taken if the control room ventilation radiation monitor is not

operable.

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 proposed change clarifies Technical Specification 3.3.7.1,

``Radiation Monitoring Instrumentation'' by revising Action 72 (for an

inoperable Control Room Ventilation Radiation Monitor) to remove

several inconsistencies between it and Action 3.7.2.b.2 of the Control

Room Emergency Recirculation System Specification. Revised Action 72

simply makes the two Specifications more consistent by incorporating

alternative compensatory measures that the operators may take after the

Control Room Ventilation Radiation Monitor has been inoperable for more

than seven days. The proposed Action would retain the choice of

initiating at least one train of the Control Room Emergency

Recirculation System, while providing a second option to take which

would depend on the current Operational Condition. In Operational

Conditions 4, 5 and * * * the current Specification 3.3.7.1 Action 72

does not contain the provisions of the Control Room Emergency

Recirculation System Action 3.7.2.b.2 which directs the Operators to

suspend performance of Core Alterations, handling of irradiated fuel

and operations with a potential for draining the reactor vessel instead

of initiating the Control Room Emergency Recirculation System. This

inconsistency between the two specifications has caused compliance

difficulties; therefore, the proposed Action adds this alternative.

Also, in Operational Conditions 1, 2 and 3 a shutdown provision is

being added. The other changes are editorial, in order to clarify the

applicability of the proposed alternative compensatory measures, to be

consistent with PNPP-specific terminology, and to be more consistent

with Action b of Specification 3.7.2.

In summary, there is no change in the probability or consequences

of any accident since the revision of Specification 3.3.7.1 Action 72

is simply proposed in order to achieve consistency with the current

Action 3.7.2.b.2. Incorporation of the already approved 3.7.2.b.2

compensatory measures to suspend possible radiation accident initiating

activities provides an alternative which would actually reduce the

probability of occurrence of a previously analyzed accident, and would

have no adverse effect on accident consequences. None of the proposed

changes to the clarified action, including the editorial changes,

involves a change to the design of the plant, nor the operational

characteristics of any plant system, nor the procedures by which the

Operators run the plant.

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

different kind of accident from any accident previously evaluated.

No design changes are being made that would create a new type of

accident or malfunction, and the methods and manner of plant operation

remains unchanged. The proposed revisions to Action 72 will remove

several inconsistencies between the two Specifications by providing

consistent actions within the Radiation Monitoring Instrumentation

Specification with those currently existing in the Control Room

Emergency Recirculation System Specification and provide an additional

shutdown requirement in Operational Conditions 1, 2 and 3. The other

changes to Action 72 are editorial, and therefore cannot affect

accident initiation parameters. The instrument to which Action 72

applies (the Control Room Ventilation Radiation Monitor (Noble Gas))

simply serves as a supporting instrumentation channel for the Control

Room Emergency Recirculation System, therefore no new or different kind

of accident can be created.

3. The proposed changes do not involve a significant reduction in a

margin of safety.

The proposed change to Specification 3.3.7.1 Action 72 simply makes

the two Specifications more consistent by making the Action for a

supporting instrumentation channel, the Control Room Ventilation

Radiation Monitor (Noble Gas), more consistent with those of the

supported system Specification, the Control Room Emergency

Recirculation System. A shutdown requirement is also being added if the

operators should choose not to initiate the supported system in

Operational Conditions 1, 2, and 3. Since the Actions of the two

Specifications will now correspond, the margin of safety as currently

exists today for the governing Specification (the Control Room

Emergency Recirculation System Specification) is maintained and the

proposed changes do not therefore reduce 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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

Attorney for Licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &

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

NRC Project Director: John N. Hannon.

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

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

Date of Amendment Request: March 18, 1994.

Description of Amendment Request: The proposed amendment would

revise TS 2.1.2 (Reactor Core), TS 2.2.1 (Reactor Protection System

Setpoints), Bases 2.1.1 and 2.1.2 (Reactor Core), Bases 2.2.1 (Reactor

Protection System Instrumentation Setpoints), TS 3.2.2 and 3.2.3 (Power

Distribution Limits), Bases 3/4.2 (Power Distribution Limits), and TS

6.9.1.7 (Administrative Controls, Core Operating Limits Report). This

amendment would remove cycle-specific limits from TS and relocate them

in the Core Operating Limits Report.

Basis for Proposed No Significant Hazards Consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below, indicating that the proposed

changes would:

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

accident previously evaluated because no accident initiators,

assumptions or probabilities are affected by the proposed relocation

of cycle-specific core operating limits to the Core Operating Limits

Report.

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

accident previously evaluated. The proposed changes do not affect

any equipment, accident conditions, or assumptions which could lead

to a significant increase in radiological consequences.

2a. Not create the possibility of a new kind of accident from

any accident previously evaluated because no new accident initiators

are introduced by these proposed changes.

2b. Not create the possibility of a different kind of accident

from any accident previously evaluated because no different accident

initiators are introduced by these proposed changes.

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

because the proposed changes only relocate cycle-specific core

operating limits to the Core Operating Limits Report; they do not

allow less conservative operating limits. The analytical methods to

be used in the determination of cycle-specific core operating limits

are previously approved by the NRC. The same margin of safety

provided in the current Technical Specifications will continue to be

maintained.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of Toledo Library,

Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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

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

NRC Project Director: John N. Hannon.

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

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

Date of Amendment Request: March 30, 1994.

Description of Amendment Request: The proposed amendment would add

a new TS Limiting Condition for Operation 3/4.4.12, Pilot Operated

Relief Valve and Block Valve, and would include associated Bases and

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, indicating that the proposed

additions and changes would:

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

accident previously evaluated because no change is being made to any

accident initiator. Automatic actuation of the PORV is not assumed

to mitigate the consequences of a design basis accident as described

in Chapter 15 of the USAR. The proposed changes will continue to

ensure the PORV and block valves are available to perform their

functions when required to do so. Therefore, it can be concluded

that the proposed changes do not involve a significant increase in

the probability of an accident previously evaluated.

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

accident previously evaluated because the proposed changes do not

invalidate accident conditions or assumptions used in evaluating the

radiological consequences of an accident.

2a. Not create the possibility of a new kind of accident from

any accident previously evaluated because the proposed changes do

not delete any function previously provided by the PORV nor has the

possibility of inadvertent opening been increased. No new types of

failures or accident initiators are introduced by the proposed

changes.

2b. Not create the possibility of a different kind of accident

from any accident previously evaluated because no new failure modes

have been defined for any plant system or component important to

safety, nor has any new limiting single failure been identified as a

result of the proposed changes. No different accident initiators or

failure mechanisms are introduced by the proposed changes.

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

because the proposed changes continue to ensure the availability of

the PORV and block valve when called upon to perform their function

and will not impact any safety analysis assumptions.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of Toledo Library,

Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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

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

NRC Project Director: John N. Hannon.

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

Electric Station, Units 1 and 2, Somervell County, Texas

Date of Amendment Request: February 14, 1994.

Brief Description of Amendments: The proposed amendment would

revise the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2

technical specifications to increase the Unit 2 boron concentration for

the refueling water storage tank (RWST) and the emergency core cooling

system (ECCS) accumulators to support Unit 2 operation with extended

fuel cycles. These changes are applicable to Unit 2 only and are

identical to those previously approved for Unit 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. The NRC staff has reviewed the licensee's analysis

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

presented below:

1. The proposed change would not increase the probability or

consequences of a previously evaluated accident.

The proposed changes are related to the boron concentration in the

RWST and ECCS accumulators. This increased concentration does not

constitute a change expected to increase the probability of a

previously evaluated accident. The means by which the proposed changes

might result in increased radiological consequences of various

accidents are discussed below.

The higher boron concentration may result in increased probability

of equipment failure following an accident due to in-containment or in-

process equipment being exposed to a more severe post-accident

environment. The general chemical properties of the slightly higher

boron concentration fluid indicates no mechanism that would result in

an appreciable increase in the component failure rate. While the

corrosive nature of the fluid will increase, this increase will be only

minimal. Thus, there is no significant increase in the consequences of

any accident due to an increase in the probability of equipment

failure.

The changes in containment spray and sump solution pH may change

the radioisotope removal and partition characteristics. While some

relevant characteristics are affected, the resulting limiting

coefficient values associated with the pH changes are bounded by the

values used in the design calculations for CPSES. Thus, no adverse

impact of the radiological consequences arising from this mechanism has

been identified.

The impact of the containment spray, with a lower pH, upon the

combustible gas production rate was also evaluated. No mechanism for

increased combustible gas production was identified.

The higher boron concentration could have an adverse impact on the

inadvertent actuation of the ECCS event. Although the timing of the

sequence of events may be affected, the departure from nucleate boiling

ratio continues to increase from its initial value throughout the

event. On the basis of its review of this event, the licensee has

identified no changes in the event probability or consequences;

however, the continued validity of this conclusion will be reconfirmed

by the licensee on a cycle-specific basis.

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

different kind of accident from any previously evaluated.

The proposed change only changes the allowable boron concentration.

No new or different accident sequences have been identified.

Furthermore, the licensee has reviewed the heat tracing requirements

and determined that there are no additional requirements resulting from

the boron concentration increase. There are no previously unconsidered

failure mechanisms.

3. The proposed change would not involve a significant reduction in

the margin of safety.

The decrease in the containment spray and sump solution pH could be

expected to result in higher airborne iodine concentrations. The

accident source terms could be impacted by variations in the iodine

spray removal and partition factors. A comparison of the coefficients

for the minimum equilibrium containment sump solution pH to those used

in the CPSES design analyses indicated that the expected coefficient

values would remain bounded by the values used in the previous

analyses. Thus, no significant reduction in the margin of safety has

been identified.

Based on this review, it appears that the standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the request for amendments involves no significant hazards

consideration.

Local Public Document Room Location: University of Texas at

Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.

Box 19497, Arlington, Texas 76019.

Attorney for Licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, NW., suite 1000, Washington, DC 20036.

NRC Project Director: Suzanne C. Black.

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

Electric Station, Units 1 and 2, Somervell County, Texas

Date of Amendment Request: February 14, 1994.

Brief Description of Amendments: The proposed amendment will revise

the Comanche Peak Steam Electric Station, Units 1 and 2, technical

specifications to be consistent with the new 10 CFR part 20.

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 a previously evaluated accident.

The proposed revisions to the liquid and gaseous effluent

release limits will not change the type or amount of effluent

released nor will there be an increase in individual or cumulative

dose. The changes will result in levels of radioactive materials in

effluents being maintained ALARA [as low as reasonably achievable]

and comply with 10 CFR 50.36a and 10 CFR 50 Appendix I. The change

to the high radiation area dose measurement distance will ensure

that high radiation areas are conservatively posted per 10 CFR

20.1601(a)(1) and provide controls to minimize individual dose. The

changes do not impact the operation or design of any plant

structure, system or component. Other proposed changes are

administrative only. Therefore, the proposed changes do not involve

an increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not affect the plant design or operation

nor do they result in a change to the configuration of any

equipment. No change is proposed that will change the type or

quantity of effluents released off site or change the source terms

available for release. Therefore, 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.

The proposed changes do not change the type or increase the

amount of effluents released offsite. No change in the methodology

used to control radioactive waste or radiological environmental

monitoring is proposed. Control of radioactive effluents and

effluent monitor setpoints will be based on current dose to the

public limitations. Under the proposed change, high radiation area

measurements are more conservative and will not result in an

increase in individual or cumulative occupational radiation

exposures. Compliance with the limits of the revised 10 CFR 20.1301

will be demonstrated by operating within the limits of 10 CFR 50,

Appendix I and 40 CFR 190. Therefore, these changes do not reduce

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of Texas at

Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.

Box 19497, Arlington, Texas 76019.

Attorney for Licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street NW., suite 1000, Washington, DC 20036.

NRC Project Director: Suzanne C. Black.

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

Electric Station, Units 1 and 2, Somervell County, Texas

Date of Amendment Request: February 14, 1994.

Brief Description of Amendments: The proposed amendment would

revise the Comanche Peak Steam Electric Station Units 1 and 2 technical

specifications by reducing the frequency of reports for radiological

effluents from semiannual to annual, and change the due date from

within 60 days after January 1 and July 1 to prior to May 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 changes do not involve a significant increase in

the probability or consequences of a previously evaluated accident.

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The amendment involves only changes of reporting

frequency and due date requirements for radiological effluent

release reporting. These changes are administrative in nature and do

not affect safe operation of the plant; therefore, accident

probabilities or consequences are unaffected.

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

or different kind of accident from any accident previously

evaluated.

The proposed amendment does not create the possibility of a new

or different kind of accident from any accident previously

evaluated. The proposed amendment is administrative in nature and

does not involve any changes to plant design of configuration. For

this reason, it will not create the possibility of a new or

different kind of accident.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The proposed amendment does not involve a significant reduction

in the margin of safety. The proposed amendment only changes the

reporting frequency and due date requirements for radiological

effluent release reporting. The reporting requirements for

radiological effluent releases are administrative changes:

therefore, there is not a significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of Texas at

Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.

Box 19497, Arlington, Texas 76019.

Attorney for Licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street NW., suite 1000, Washington, DC 20036.

NRC Project Director: Suzanne C. Black.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of Amendment Request: March 30, 1994.

Description of Amendment Request: The proposed changes would revise

the North Anna Power Station, Units No. 1 and No. 2 (NA-1&2) Technical

Specifications (TS). Specifically, the proposed changes would revise

the High Head Safety Injection (HHSI) flow balance surveillance

requirements by removing specific numerical values. The numerical

values would be replaced with broader requirements to ensure that the

HHSI flow rates meet the loss of coolant accident (LOCA) analysis

acceptance criteria and pump runout limits. The NA-1&2 TS 4.5.2.h

requires a surveillance test of the HHSI system following the

completion of any modification to the Emergency Core Cooling System

(ECCS) subsystems that could alter the subsystem flow characteristics.

The current surveillance criteria specify values for the sum of the

injection line flow rates, excluding the highest flow rate, and the

total pump flow rate. These correspond to requirements for the safety

analysis flow input and the HHSI pump runout limit, respectively.

The HHSI test acceptance criteria in the current TS are very narrow

because of the various system physical and technical constraints that

need to be considered in the flow balance testing. These acceptance

criteria may also be more restrictive than required by either the LOCA

analysis or the actual pump runout requirements. For example, the LOCA

analysis contains input conservatisms that could be used to offset a

reduction in the required HHSI flow while still meeting the 10 CFR

50.46 LOCA acceptance criteria. The proposed TS changes would permit

the use of additional available margin, while maintaining a strong

technical linkage between the measured system performance and the

safety analysis. Although these proposed TS changes remove the

numerical values from TS 4.5.2.h, neither the methodology nor the

acceptance criteria for LOCA analysis are affected.

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:

Specifically, operation of North Anna Power Station in

accordance with the proposed Technical Specification changes will

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. The proposed

Technical Specification changes continue to require that with one

HHSI pump running, the sum of the flows through the two lowest

branch lines shall be [greater than or equal to] the minimum HHSI

flow required by the safety analysis and that the total HHSI pump

flow rate shall be [less than or equal to] the evaluated HHSI pump

runout limit.

Likewise, the consequences of the accidents previously evaluated

will not increase as a result of the proposed Technical

Specification changes. The system performance will remain bounded by

the safety analysis for all postulated conditions. The safety

analysis will continue to be performed and evaluated in accordance

with the requirements of 10 CFR 50.59 and 10 CFR 50.46.

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

or malfunction from any previously evaluated. The proposed Technical

Specification changes will not affect the capability of the HHSI

System to perform its intended function. The proposed Technical

Specification changes are bounded by the existing safety analysis

and do not involve operation of plant equipment in a different

manner from which it was designed to operate.

Since a new failure mode is not created, a new or different type

of accident or malfunction is not created.

3. Involve a reduction in a margin of safety. The system

performance will continue to bound the flow rates specified in the

safety analysis, therefore safety margins are not reduced.

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: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for Licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: Herbert N. Berkow.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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

County, Illinois

Date of Application for Amendments: March 21, 1994.

Description of Amendment Requests: The proposed amendments would

permit continued activities at all four units with main steam Code

safety valve tolerances of plus or minus 3% until the lift setpoints

can be reset to within plus or minus 1%.

Date of Publication of Individual Notice in Federal Register: March

29, 1994 (59 FR 14685).

Expiration Date of Individual Notice: April 29, 1994.

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.

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

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

Date of Amendment Request: March 24, 1994.

Description of Amendment Request: The proposed amendment would

revise section 6.0 (Administrative Controls). Specifically, the plant

staff requirement (specified in Technical Specification (TS) 6.2.2.i)

would be revised to temporarily allow the operations manager to have

held a senior reactor operator (SRO) license at a pressurized water

reactor (PWR) other than Indian Point 3. The TS currently requires the

operations manager to have or have held an SRO license at Indian Point

3 only. This proposed change is needed to allow management changes at

the facility in an effort to improve overall performance. The proposed

changes would be in effect for a period ending 3 years after restart

from the 1993/1994 Performance Improvement Outage.

Date of Publication of Individual Notice in Federal Register: April

1, 1994 (59 FR 15464).

Expiration Date of Individual Notice: May 3, 1994.

Local Public Document Room Location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for Licensee: Charles M. Pratt, 10 Columbus Circle, New

York, New York 10019.

NRC Project Director: Robert A. Capra

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.

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

529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,

and 3, Maricopa County, Arizona

Date of Application for Amendments: December 2, 1993.

Brief Description of Amendments: The amendments will modify TS 3/

4.6.1.2 by removing the schedular requirements for a Type A (overall

integrated containment leakage rate) test to be performed specifically

at 40 plus or minus 10-month intervals and replacing these requirements

with a requirement to perform Type A testing in accordance with

Appendix J to 10 CFR part 50.

Date of Issuance: April 6, 1994.

Effective date: April 6, 1994.

Amendment Nos.: 73, 59, and 45.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the Technical Specifications.

Date of Initial Notice in Federal Register: January 5, 1994 (59 FR

616) The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 6, 1994.

No significant Hazards Consideration Comments Received: No.

Local Public Document Room Location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,

Maryland

Date of Application for Amendments: August 27, 1993, as

supplemented March 11, 1994.

Brief Description of Amendments: The amendments revise the Calvert

Cliffs Nuclear Power Plant, Units 1 and 2, Technical Specifications

(TSs) by removing the list of containment isolation valves in Table

3.6-1. The amendments also make accompanying changes to various TSs and

to the TS Bases. These amendments are a ``line-item'' TS improvement

and follow the guidance of Generic Letter 91-08, ``Removal of Component

Lists From Technical Specifications.''

Date of issuance: April 7, 1994.

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

30 days.

Amendment Nos.: 187 and 164

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Technical Specifications.

Date of Initial Notice in Federal Register: September 29, 1993 (58

FR 50966) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated April 7, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

Station, Plymouth County, Massachusetts

Date of Application for Amendment: June 7, 1993, August 9, and

December 10, 1993.

Brief Description of Amendment: This amendment revises the

Technical Specification (TS) to support a 24-month fuel cycle. The TS

changes include extending surveillance intervals and adjusting

setpoints as justified in the Safety Evaluation.

Date of Issuance: April 6, 1994.

Effective Date: April 6, 1994.

Amendment No.: 151.

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

Technical Specifications.

Date of Initial Notice in Federal Register: March 16, 1994 (59 FR

2863) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 6, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

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

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

Date of Application for Amendments: March 26, 1993.

Brief Description of Amendments: The amendments modify the trip

level settings for the Isolation Condenser and High Pressure Core

Injection System Steam lines to more conservative values. In addition,

the proposed amendments revise the ECCS Low-Low Water Level initiation

trip setting to a more conservative number.

Date of Issuance: April 5, 1994.

Effective Date: April 5, 1994.

Amendment Nos.: 126 and 120.

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

revised the Technical Specifications.

Date of Initial Notice in Federal Register: March 2, 1994 (59 FR

10002) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 5, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Morris Public Library, 604

Liberty Street, Morris, Illinois 60450.

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

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

Date of Application for Amendments: January 24, 1994.

Brief Description of Amendments: The amendments implement line item

5.9 of Generic Letter 93-05, ``Line-Item Technical Specifications

Improvements to Reduce Surveillance Requirements for Testing During

Power Operation'', which provided recommendations for deleting the

requirement to perform response time testing where the required time

corresponds to the diesel start time.

Date of issuance: April 7, 1994.

Effective Date: April 7, 1994.

Amendment Nos.: 98 and 82.

Facility Operating License Nos. NPF-11 and NPF-18. The amendments

revised the Technical Specifications.

Date of Initial Notice in Federal Register: February 16, 1994 (59

FR 7686). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 7, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Public Library of Illinois

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

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

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

Date of Application for Amendments: February 22, 1993 as

supplemented August 16, 1993.

Brief Description of Amendments: The amendments allow continued

operation of one unit for a period of seven days while the common plant

(Division 1) emergency diesel generator (``O'' DG) is out of service

for the performance of specified Technical Specification surveillance

requirements and the performance of planned maintenance and/or

modification work. Also, the amendments clarify Surveillance

Requirement 4.8.1.1.2.a.7 to allow an emergency diesel generator to

remain Operable with only one air start subsystem pressurized.

Date of Issuance: April 11, 1994.

Effective Date: April 11, 1994.

Amendment Nos.: 99 and 83.

Facility Operating License Nos. NPF-11 and NPF-18. The amendments

revised the Technical Specifications.

Date of Initial Notice in Federal Register: July 7, 1993 (58 FR

36430) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 11, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Public Library of Illinois

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

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

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

Date of Application for Amendments: September 28, 1993, as

supplemented February 17, 1994.

Brief Description of Amendments: The amendments delete the portion

of the 18-month surveillance requirement contained in Technical

Specification (TS) 4.5.2.d associated with verifying that the decay

heat removal system suction isolation valves automatically close on a

reactor coolant system pressure signal. Also, an obsolete footnote to

TS 4.5.2.e is being deleted. This footnote is no longer necessary since

the first Unit 1 refueling outage is complete.

Date of Issuance: April 4, 1994.

Effective Date: April 4, 1994.

Amendment Nos.: 117 and 111.

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

revised the Technical Specifications.

Date of Initial Notice in Federal Register: March 2, 1994 (59 FR

10004) The February 17, 1994, letter provided clarifying information

that did not change the scope of the initial September 28, 1993,

application and initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 4, 1994.

No significant hazards consideration comments received: No.

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

Black Street, Rock Hill, South Carolina 29730.

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

No. 2, Pope County, Arkansas

Date of Application for Amendment: July 22, 1993, as supplemented

by letter dated October 20, 1993.

Brief Description of Amendment: The amendment removed the cycle-

specific variables from the Technical Specifications (TSs) and

controlled them under a new document called the Core Operating Limits

Report (COLR), in accordance with Generic Letter 88-16.

Date of Issuance: April 11, 1994.

Effective Date: April 11, 1994.

Amendment No.: 157.

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

Technical Specifications.

Date of Initial Notice in Federal Register: September 1, 1993 (58

FR 46230). The additional information contained in the supplemental

letter dated October 20, 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 April 11, 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: January 13, 1994.

Brief Description of Amendment: The amendment requested the removal

of the temporary technical specification limit on the number of spent

fuel assemblies that may be stored in the spent fuel pool at Grand Gulf

Nuclear Station pending licensee verification of the adequacy of the

spent fuel pool heat removal capability.

Date of Issuance: April 4, 1994.

Effective Date: April 4, 1994.

Amendment No: 113.

Facility Operating License No. NPF-29. Amendment revises the

Technical Specifications.

Date of Initial Notice in Federal Register: March 2, 1994 (59 FR

10006) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 4, 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.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County, Florida

Date of Application for Amendments: April 20, 1993.

Brief Description of Amendments: These amendments delete the lead/

lag compensator term on the measured reactor coolant system loop

temperature difference from the overtemperature and overpower Delta T

reactor trip functions.

Date of Issuance: April 4, 1994.

Effective Date: April 4, 1994.

Amendment Nos. 161 and 155.

Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments

revised the Technical Specifications.

Date of Initial Notice in Federal Register: June 9, 1993 (58 FR

32383) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 4, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Florida International

University, University Park, Miami, Florida 33199.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and

50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,

Georgia

Date of Application for Amendments: September 20, 1993.

Brief Description of Amendments: The amendments revise the Units 1

and 2 Channel Functional Test frequency from quarterly to once per 18

months for the scram discharge volume float type level switches.

Date of Issuance: April 15, 1994.

Effective Date: To be implemented within 60 days from the date of

issuance.

Amendment Nos.: 193 and 133.

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

revised the Technical Specifications.

Date of Initial Notice in Federal Register: October 27, 1993 (58 FR

57852) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 15, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513.

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas, Docket

No. 50-499, South Texas Project, Unit 2, Matagorda County, Texas

Date of Amendment Request: January 25, 1994.

Brief Description of Amendment: The amendment added new Technical

Specifications, 3/4.10.6 and 3/4.10.7, to the Special Test Exceptions

section. TS 3/4.10.6 allows the restart of Unit 2 with expired

calibrations on the core exit thermocouples (CET) and the reactor

coolant system (RCS) resistance temperature detectors (RTD) by setting

aside the affected limiting conditions for operation (LCOs) until the

calibrations are complete. This is a one-time only change that is valid

during the third refueling outage for Unit 2 until the calibrations are

complete. TS 3/4.10.7 adds a new technical specification to allow the

ascension to 75 percent rated thermal power with an expired precision

heat balance reactor coolant flow measurement. This change is effective

only for Unit 2, Cycle 4, until the surveillance requirement is

completed.

Date of Issuance: April 1, 1994.

Effective Date: April 1, 1994, to be implemented within 10 days of

issuance.

Amendment No.: Amendment No. 48.

Facility Operating License No. NPF-80. Amendment revised the

Technical Specifications.

Date of Initial Notice in Federal Register: February 16, 1994 (59

FR 7690) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 1, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.

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

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

Date of Application for Amendment: January 21, 1994.

Brief Description of Amendment: The amendment revises Technical

Specification 4.6.3 (Emergency Power Sources), to eliminate unnecessary

testing of an operable emergency diesel generator (EDG) when the

redundant EDG becomes inoperable. This amendment is intended to

increase EDG reliability and the overall level of plant safety by

reducing the stresses on the EDGs caused by unnecessary testing. This

amendment also eliminates the requirement to load the operable EDG with

the offsite network when it is tested with one EDG inoperable.

Date of Issuance: April 6, 1994.

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

30 days.

Amendment No.: 147.

Facility Operating License No. DPR-63: Amendment revises the

Technical Specifications.

Date of Initial Notice in Federal Register: March 2, 1994 (59 FR

10009) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 6, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Reference and Documents

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

York 13126.

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

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

Date of Application for Amendment: November 18, 1993.

Brief Description of Amendment: The amendment revises the setpoints

for the degraded volta

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