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 18, 1994, through April 29, 1994. The

last biweekly notice was published on April 28, 1994 (59 FR 22000).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

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

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to Room 6D22, Two White Flint North, 11555 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

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

20555. The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By June 10, 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.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of amendment request: March 30, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.4.9, Pressure/Temperature

Limits, and its associated Bases, by changing the Unit 1 heatup and

cooldown curves to incorporate a newly determined reactor vessel

reference nil-ductility temperature, RTNDT, and by updating the

removal schedule of vessel surveillance capsules for both units in

accordance with ASTM E185-82. Changes would also be made to the Unit 1

Low Temperature Overpressure Protection System (LTOPS) setpoint curve

in TS 3.4.9.3 to reflect the new pressure/temperature limits.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

A. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The use of new pressure-temperature limit curves and low

temperature overpressure protection curves will not change any

postulated accident scenarios. The revised curves were developed using

industry standards and regulations which are recognized as being

inherently conservative. The pressure-temperature low temperature

overpressure curves provide reactor coolant system (RCS) limits to

protect the reactor pressure vessel from brittle fracture by clearly

separating the region of normal operations from the region where the

vessel is subject to brittle fracture. The heatup and cooldown limits

are designed to ensure that the 10 CFR 50 Appendix G Pressure

Temperature limits for the RCS are not exceeded during any condition of

normal operation including anticipated operational occurrences.

General Design Criterion 32 of 10 CFR 50 Appendix A requires that

the reactor coolant boundary shall be designed with sufficient margin

to assure that when stressed under operating, maintenance, testing, and

postulated accident condition[s], (1) the boundary behaves in a

nonbrittle manner and (2) the probability of rapidly propagating

fracture is minimized.

10 CFR 50 Appendix G, ``Fracture Toughness Requirements,'' requires

that the effects of changes in the fracture toughness of reactor vessel

materials caused by neutron radiation throughout the service life of

[a] nuclear reactor be considered in the pressure-temperature limits.

The change is used in conjunction with the material initial reference

temperature (RTNDT) to establish the limiting pressure-temperature

curves. Regulatory Guide 1.99, Rev. 2, contains procedures for

calculating the effects of neutron radiation embrittlement of the low-

alloy steels currently used for light-water-cooled reactor vessels.

Using the Regulatory Guide 1.99, Revision 2, Braidwood Unit 1

Surveillance Capsule U results, and Appendix G to 10 CFR 50, new

Pressure-Temperature curves [were] prepared for the projected reactor

vessel exposure at 32 EFPY of operation. These new curves, in

conjunction with the heatup and cooldown ranges and the revised Low-

Temperature Overpressure Protection System setpoints, provide the

required assurance that the reactor pressure vessel is protected from

brittle fracture up 32 EFPY of operation. No changes to the design of

the facility have been made and no new equipment has been added or

removed. The revised analysis and resultant adjustment of the operating

limitations provide assurance that the Reactor Coolant System is

protected from brittle fracture.

Revising the Reactor Vessel Material Surveillance Program

Withdrawal Schedule does not result in the addition or removal of any

equipment, or any design changes to the facility. Capsule lead times

are revised and, for Braidwood Unit 2, Capsule X will be removed next

vice Capsule W. The proposed removal schedules remain consistent with

ASTM 185-82.

Therefore, the proposed amendment to the pressure temperature

limitations does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

The use of the new pressure-temperature operating limits and the

new low temperature overpressure protection curve does not change any

postulated accident scenarios. The new curves do not represent any

appreciable change in the current methodologies; they merely provide

assurance that the Reactor Coolant System is protected from brittle

fracture. No new accident or malfunction mechanism is introduced by the

amendment and no physical plant changes will result from this

amendment.

Revision of the Reactor Vessel Material Surveillance Program

Withdrawal Schedule does not introduce a new accident or malfunction

mechanism. Capsule lead times are revised, and, other than changing the

order of specimen removal, consistent with ASME 185-82, no physical

plant changes will result from this revised schedule.

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

new or different kind of accident from any accident previously

evaluated.

C. The proposed change does not involve a significant reduction in

a margin of safety.

The new pressure-temperature operating limits low temperature

overpressure protection curves were generated with the currently

accepted conservative methodology using capsule surveillance data. The

new pressure-temperature curves were developed using industry standards

and regulations (ASME Code Section III, and NRC Regulatory Guide 1.99,

Revision 2) which are recognized as being inherently conservative. The

use of the new pressure- temperature operating limits and low

temperature overpressure protection limits would not change postulated

accident scenarios.

The proposed revision to the Reactor Vessel Material Surveillance

Program Withdrawal Schedule would not change postulated accident

scenarios. Capsule lead times are revised, and, other than changing the

order of specimen removal, consistent with ASTM 185- 82, no physical

plant changes will result from this revised schedule. 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: Wilmington Township Public

Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

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

Austin, One First National Plaza, Chicago, Illinois 60690.

NRC Project Director: James E. Dyer.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of amendment request: March 31, 1994.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TS) to provide allowable outage

times for automatic actuation channel surveillance testing and

restoration time for an inoperable engineered safety feature actuation

system automatic actuation channel.

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 following evaluation is provided for the three categories of

the significant hazards consideration standards:

a. Proposed changes to allow 8 hours for master relay and logic

testing, 12 hours for slave relay testing and 6 hours to restore an

inoperable ESFAS Automatic Actuation Channel prior to entering the

shutdown action clock.

(1) The determination that these changes are within all acceptable

criteria was established in the NRC's SER prepared for WCAP-10271,

Supplement 2, Revision 1. The Technical Specification changes proposed

by this license amendment request conform to NRC guidance contained in

the SER. The NRC found that implementation of the proposed changes is

expected to result in a small and acceptable increase in ESFAS

unavailability. This increase in probability results in a small

increase in calculated core damage frequency and public risk. The

calculated increase in core damage frequency was judged to be

acceptable since the increase was small and well within the range of

uncertainty associated with the analysis. The values presented in WCAP-

10271 Supplement 2 Revision 1 for increase in core damage frequency

were verified by Brookhaven National Laboratory as part of an audit and

sensitivity analyses performed for the NRC Staff.

Based on the small increase in core damage frequency as compared

with the range of uncertainty in the analysis, the NRC agreed that the

calculated increase is acceptable. This conclusion was documented in

the NRC's SER dated February 22, 1989. The applicability of these

conclusions has been verified through a plant specific review of the

generic analysis in WCAP-10271, Supplement 2, Revision 1. The ESFAS

Automatic Actuation Channel allowed outage and restoration times

included in this license amendment request are consistent with the

generic analysis. In addition, the NRC stated that the majority of the

increase in unavailability was due to the decrease in frequency of

surveillance testing vice the changes in allowed outage and restoration

times. Therefore, considering the above information, the proposed

allowed outage and restoration time changes do not involve a

significant increase in the probability of occurrence or consequences

of an accident previously evaluated.

(2) The proposed changes do not involve the physical alteration of

any plant system and do not result in a change in the manner in which

the ESFAS system performs its function. The increases in allowed outage

and restoration times only affects the probability of the ESFAS

Automatic Actuation Channel functioning properly as described above.

Therefore, the allowed outage and restoration time changes proposed in

this license amendment request do not create a new or different type of

accident from any previously evaluated.

(3) The proposed allowed outage time and restoration time changes

do not alter the manner in which safety limits, limiting safety system

setpoints or limiting conditions for operation are determined. The

impact of the revised ESFAS Automatic Actuation Channel allowed outage

and restoration times is addressed above. Implementation of the

proposed changes is expected to result in an overall improvement in

safety by allowing adequate time for required ESFAS testing and quality

repairs leading to improved equipment reliability due to a more

appropriate restoration time. Therefore, it may be concluded that the

proposed allowed outage and restoration time changes do not involve a

significant reduction in margin of safety.

b. Proposed change to the minimum required degree of redundancy for

the High-High Containment Pressure channels in Table 3.4-1.

(1) Changing the minimum required degree of redundancy in Table

3.4-1 for the High-High Containment Pressure Channels (Table 3.4-1

items II.3, III.B.3, and IV.3) provides consistency with Technical

Specification 3.4.2.c which allows an inoperable High-High Containment

Pressure channel to be placed in bypass. Placement of an inoperable

High-High Containment Pressure Channel in bypass is preferred to reduce

the probability of an inadvertent containment spray event. Also, these

channels are designed with a two out of four logic so that the failed

channel may be bypassed rather than tripped. With the failed channel

bypassed, single failure criterion is still met because the logic is

now a two out of three. Furthermore, with the one channel bypassed, a

single channel failure will not inadvertently initiate a containment

spray. Therefore, this change can be considered an administrative

change to correct Table 3.4-1 to agree with the Action requirements of

Technical Specification 3.4.2.c. As such this proposed change does not

involve an increase in the probability of occurrence or consequences of

an accident previously evaluated.

(2) Correcting the minimum required degree of redundancy in Table

3.4-1 for the High-High Containment Pressure channels is an

administrative change which does not involve the physical alteration of

any plant system and does not result in a change in the manner in which

the ESFAS system performs its function. Therefore, the proposed

correction to Table 3.4-1 does not create the possibility of a new or

different kind of accident from any previously analyzed.

(3) Correcting the minimum required degree of redundancy in Table

3.4-1 to be consistent with the Actions of Technical Specification

3.4.2.c is an administrative change and as such does not involve any

reduction in a margin of safety.

c. Proposed change to the delete footnote +++ from Table 3.4-1.

(1) Deleting footnote +++ from Table 3.4-1 removes the

inconsistency between it and Technical Specification 3.4.2.c which

states that channels other than the High-High Containment Pressure

channels shall be placed in trip during testing. The change does not

affect the manner in which ESFAS provides plant protection. In addition

the change does not affect the functioning of ESFAS or the way Zion

Station conducts channel testing. Instrument channel testing will

continue to be conducted in the tripped mode with the exception of the

High-High Containment Pressure channels, which can be tested in bypass

because of the risk of a spurious Containment Spray event. Automatic

Actuation Channel testing will be performed in accordance with the

allowed outage times of new Specification 3.4.2.d. As such this

proposed change does not involve any significant increase in the

probability of occurrence or consequences of an accident previously

evaluated.

(2) Deleting footnote +++ from Table 3.4-1 does not involve the

physical alteration of any plant system and does not result in a change

in the manner in which ESFAS performs its function. Therefore this

change does not involve the physical alteration of any plant system and

does not result in a change in the manner in which the ESFAS system

performs its function. Therefore, the proposed correction to Table 3.4-

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

accident from any previously analyzed.

(3) Deleting footnote +++ from Table 3.4-1 does not alter the

manner in which safety limits, limiting safety system setpoints or

limiting conditions for operation are determined. Implementation of

this change will not alter ESFAS testing. Therefore implementation of

this change does not involve any reduction in a margin of safety.

d. Proposed editorial change to Technical Specification 3.4.2.c.

The editorial change to Technical Specification 3.4.2.c to change

``Containment Hi-Hi pressure channels'' to ``High-High Containment

Pressure channels'' is purely an administrative change which has no

affect on plant safety.

e. Summary.

The foregoing analyses demonstrate that the proposed License

Amendment to the Zion Station Technical Specifications does not involve

a significant increase in the probability of occurrence or consequences

of a previously evaluated accident, does not create the possibility of

a new or different kind of accident and 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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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

Austin, One First National Plaza, Chicago, Illinois 60690.

NRC Project Director: James E. Dyer.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of amendment request: April 22, 1994.

Description of amendment request: The proposed change revises the

reactor vessel pressure-temperature limits in the Technical

Specifications. The change insures that the vessel fracture toughness

requirements of Section V of 10 CFR Part 50, Appendix G, are satisfied

through the end of life.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will the proposed change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The margin above Nil Ductility Transition Temperature (NDTT) is

governed by 10 CFR 50 Appendix G and remains unchanged. The proposed

change will not involve a significant increase in the probability or

consequences of a previously evaluated accident.

2. Will the proposed change create the possibility of a new or

different kind of accident from any accident previously evaluated?

The predicted shifts in NDTT are based on a revised reference

temperature consistent with Regulatory Guide 1.99, Revision 2, dated

May 1988. This method of revising temperature-pressure limits is the

same as in the past (ASME Code Section III, Appendix G).

3. Will the proposed change involve a significant reduction in the

margin of safety?

The proposed curves were generated for an End of Licensed Life (May

31, 2000) Effective Full Power Year exposure and are conservative in

nature until that time. The margin of safety [is] unchanged.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770

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

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Project Director: L. B. Marsh.

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

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

Date of amendment request: April 18, 1994.

Description of amendment request: License Amendment No. 81, issued

on July 15, 1993, changed the numbering of surveillance requirements

for Technical Specifications 3/4.3.1, ``Control Rod Operability,'' 3/

4.3.2, ``Control Rod Maximum Scram,'' and 3/4.10.2, ``Rod Pattern

Control System.'' However, Action Statements referencing these

surveillance requirements were overlooked and were not appropriately

renumbered. The purpose of the proposed technical specification change

would be to renumber the overlooked references.

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) These changes do not affect the intent or implementation of the

applicable Technical Specifications. The changes simply make the

affected Technical Specifications consistent. Since these are only

editorial changes which do not impact the plant design or operations,

they cannot increase the probability or the consequences of any

accident previously evaluated.

(2) The proposed changes are editorial only and do not affect the

plant design or operation. No new failure modes are introduced by such

changes and, therefore, the request will not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

(3) The proposed changes merely correct an editorial oversight.

These changes do not alter or delete any technical requirements and,

therefore, do 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: Vespasian Warner Public

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

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

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

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: April 18, 1994.

Description of amendment request: Test methods for carbon adsorber

filters specified in Technical Specification Sections 3/4.6.6.3,

``Standby Gas Treatment System,'' and 3/4.7.2, ``Control Room

Ventilation System,'' specify the 1979 version of ASTM D3803. The

proposed change would delete the year of the standard so that more

recent versions of the standard could be used.

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 to the Technical Specification

surveillance requirements for determining the methyl iodide penetration

of carbon samples would not involve a significant increase in the

probability or the consequences of any accident previously evaluated

because the proposed change merely allows Illinois Power (IP) to

utilize a more up-to-date version of the same test method currently

specified. More recent versions of the test method are more effective

at detecting unsatisfactory charcoal performance because they include

equilibration periods to ensure that all samples have a common starting

point before being challenged with radioactive gas. The proposed change

would not affect the quality of the charcoal or the reliability of the

filter subsystems as it only relates to testing and involves no changes

to the design or operation of the ventilation subsystems themselves.

The updated standards provide more accurate and repeatable test results

and do not change the properties or acceptance criteria for these

properties. As a result, the performance capabilities of the associated

filter subsystems would not be adversely impacted by the proposed

change.

(2) The proposed change would not involve a change in the design or

operation of any plant system or component. In addition, the proposed

change would not reduce the level of filter train subsystem reliability

nor would it create an initiating event for any accident. Because the

performance, function, and redundancy of the original design remain

unchanged, the proposed change would not create the potential for a new

event. Furthermore, since no new types of equipment would be introduced

into the plant design and the proposed change would not adversely

impact existing equipment, no potential for a different type of

malfunction is created by the proposed change. Therefore, this proposed

change cannot create the possibility of a new or different kind of

accident from any accident previously evaluated.

(3) The margin of safety for the charcoal filter subsystems as

defined in the Bases to the Technical Specifications associated with

the proposed change refers to the ability of the filters to remove

radioiodines. The proposed change would allow IP to upgrade the

currently specified test for determining charcoal adsorber performance

with one which utilizes the same type of methodology, but provides

greater accuracy and repeatability. The newer versions of the test

method are more effective at detecting unsatisfactory charcoal

performance because they include equilibration periods to ensure that

all samples have a common starting point before being challenged with

radioactive gas. Thus, the proposed change would not involve a

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

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

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

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

NRC Project Director: John N. Hannon.

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

Generating Plant, Wright County, Minnesota

Date of amendment request: March 28, 1994.

Description of amendment request: The proposed amendment would

revise technical specifications Tables 3.2.4 and 4.2.1, to change one

of the initiating parameters of the reactor building ventilation

isolation system and standby gas treatment system (SGTS) from Low

Reactor Water Level to Low Low Reactor Water Level. This revision is

being made in order to improve plant performance by reducing the

potential for unnecessary secondary containment isolation and SGTS

initiations.

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 amendment will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The function of the Standby Gas Treatment System and secondary

containment is to mitigate the consequences of a loss of coolant

accident and fuel handling accidents. The proposed changes maintain

this capability. The revised Standby Gas Treatment System initiation

and secondary containment isolation parameter of low low reactor water

level provides the required detection of loss of coolant accidents and

is consistent [with] ECCS actuation to mitigate the consequences of

this accident. The low low reactor water level instrumentation is set

to trip when reactor water level is 6'6'' above the top of the active

fuel. This trip currently initiates closure of the Group 1 Primary

containment isolation valves, activates the Emergency Core Cooling

systems and starts the emergency diesel generator. This trip setting

level was chosen to be low enough to prevent spurious operation but

high enough to initiate Emergency Core Cooling system operation and

primary system isolation so that no melting of the fuel cladding will

occur, post accident cooling can be accomplished, and the guidelines of

the 10 CFR 100 will not be violated. Therefore, this amendment will not

cause a significant increase in the probability or consequences of an

accident previously evaluated for the Monticello plant.

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

different kind of accident from any accident previously analyzed. The

proposed changes to Technical Specifications for the standby gas

treatment system and secondary containment do not alter the function 8

of the systems or its interrelationships with other systems. An adverse

interaction which could be postulated to occur is the initiation of the

Standby Gas Treatment System without a coordinated trip of the

Mechanical Vacuum Pump. The Mechanical Vacuum Pump is operated during

plant startups to draw a vacuum on the main condenser prior to

admission of steam. The Mechanical vacuum discharges to the offgas

stack and thus can create a back pressure on the Standby Gas Treatment

System, reducing initiation of Standby Gas Treatment System flow below

required values. The proposed initiation of Standby Gas Treatment

System on low low reactor water level maintains the necessary

coordination by having the Standby Gas Treatment System initiate

subsequent to isolation or tripping of the Mechanical Vacuum Pump on a

low reactor water level signal from the primary containment isolation

logic. Therefore, this amendment will not create the possibility of a

new or different kind of accident from any accident previously

analyzed.

The proposed amendment will not involve a significant reduction in

the margin of safety.

The proposed amendment changes the initiation of the Standby Gas

Treatment System and secondary containment isolation from being

concurrent with the low reactor water signal (which is indicative that

the reactor core is in danger of being inadequately cooled) to being

concurrent with reactor low low water level (which is also an indicator

that the capability to cool the core is threatened and assures that no

melting of the fuel cladding will occur, post accident cooling can be

accomplished, and the guidelines of 10 CFR 100 will not be violated). A

review of the accident analyses provided in Section 14 of the USAR has

determined that these analyses did not specifically credit initiation

of the Standby Gas Treatment Systems and secondary containment

isolation at the accident precursor reactor water level of low level.

Furthermore, this review determined that the low low reactor water

level setpoint has no adverse impact on the ability of the Standby Gas

treatment System and secondary containment to perform its design basis

function as credited in the accident analyses.

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, Docket Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: March 24, 1994.

Description of amendment request: The proposed modification to

Technical Specification (TS) Sections 3.11.1.4, 6.9.1.8, and 6.14.1

would change the frequency for submitting the Semiannual Radioactive

Effluent Release Report to the NRC from semiannually to annually.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed TS changes are administrative in nature. The proposed

changes simply involve revising the frequency for submitting the

Semiannual Radioactive Effluent Release Report to the NRC from

semiannually to annually in order to implement the amended reporting

requirements of 10 CFR 50.36a as promulgated in Final Rule 57 FR 39353.

Since the information contained in this report is reviewed and

evaluated after the effluents are released, no accidents previously

evaluated are impacted by the proposed TS changes. Radiological

effluent releases from the station will continue to be controlled as

required by the TS, including those requirements specified in the

Offsite Dose Calculation Manual (ODCM) and Process Control Program

(PCP). The proposed TS changes do not involve any changes to the

operation or physical configuration of any plant systems or equipment.

The proposed changes do not impact any initial or final accident

conditions or assumptions previously evaluated. The radiological

consequences of these previously evaluated accidents are not affected

by the proposed changes.

Therefore, the proposed TS changes do not involve an increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The proposed TS changes are administrative in nature. The proposed

changes simply involve revising the frequency for submitting the

Semiannual Radioactive Effluent Release Report to the NRC from

semiannually to annually in order to implement the amended reporting

requirements of 10 CFR 50.36a as promulgated in Final Rule 57 FR 39353.

Radiological effluent releases from the station will continue to be

controlled as required by the TS, including those requirements

specified in the ODCM and PCP. The proposed TS changes do not involve

any modifications to plant systems or equipment.

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

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

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

in a margin of safety.

The proposed TS changes are administrative in nature, and will only

involve revising the frequency for submitting the Semiannual

Radioactive Effluent Release Report to the NRC from semiannually to

annually as currently stipulated in 10 CFR 50.36a. The specific

radiological effluent release information contained in this report will

continue to be provided as required. The station radiological effluent

releases will continue to be controlled as required by TS, including

those requirements specified in the ODCM and PCP.

Therefore, the proposed TS changes do 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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.

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: February 25, 1994.

Description of amendment request: The proposed Technical

Specification (TS) changes would permit the submittal of the

Radioactive Effluents Release Report on an annual rather than a

semiannual basis; allow changes to the Offsite Dose Calculation Manual

(ODCM) to be submitted in the Radioactive Effluent Release Report

rather than in the monthly operating report; remove the title of

Executive Vice President--Operations from the TS; remove the list of

audit frequencies from the TS and place them under Quality Systems

management; change the title of Associate Manager, Health Physics to

Radiation Protection Manager; remove references to specific letters;

remove TS 6.4 on training; and correct various typographical errors.

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

because the administrative change does not affect plant operations in

any manner.

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

or different kind of accident than previously evaluated because the

proposed change is administrative in nature and no physical alterations

of plant configuration or changes to setpoints or operating parameters

are proposed.

3. The proposed license amendment does not involve a significant

reduction in a margin of safety. The change is only administrative.

The NRC staff has reviewed the licensee's analysis and 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

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.

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: March 11, 1994.

Description of amendment request: The proposed Technical

Specification (TS) changes would delete surveillance requirement

4.8.1.4.a.3, which requires periodic testing of penetration protection

fuses, and its associated Basis.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The probability or consequences of an accident previously

evaluated is not significantly increased.

Fuses are simple protection devices and can only degrade by being

more resistive which is in the conservative direction. The proper type

and size fuse is assured as part of design, procurement, and initial

installation. The testing provides no additional assurance of

operability. Therefore, the deletion of periodic retesting of these

fuses will not increase the probability or consequences of an accident

previously evaluated.

2. [The proposed amendment will not] [c]reate the possibility of a

new or different kind of accident from any previously analyzed.

The design of the penetration protection and the installation of

the fuses has not changed in any way. Any undetected failure of a fuse

would fall under single failure criteria. A current limiting fuse must

have high electrical current in order to perform its intended function.

Any fuse which has opened the circuit through the penetration would be

detected. (This is not a concern of the Technical Specifications.)

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

different kind of accident from any previously evaluated.

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

of this surveillance requirement will not minimize the intent of this

Technical Specification. This TS is to assure continued operability of

the containment penetration conductor overcurrent protection which

helps to ensure containment integrity. Testing, however, may introduce

the potential for damage to the fuses and fuse clips. Therefore, the

deletion of this TS requirement will not involve a significant

reduction in a margin of safety.

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

this review it appears that the 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: November 22, 1993.

Description of amendment request: The proposed amendment would

result in the replacement of most of the analog Riley temperature

instrumentation associated with leak detection with digital equipment

from the General Electric Company NUMAC product line. Technical

Specification changes would be made to instrumentation surveillance

requirements for temperature instruments associated with main steam

line isolation, reactor water cleanup system isolation, reactor core

isolation cooling system isolation, and residual heat removal system

isolation.

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 Technical specifications are proposed to be revised to perform

a Channel Functional Test on a semiannual frequency versus the current

monthly frequency for both ambient and differential temperature and for

the MSL tunnel temperature timer functions, for the above listed piping

lines. Additionally, this evaluation addressees the potential for, and

implications of, common mode failures due to software, hardware and/or

electromagnetic and radio-frequency interference (EMI/RFI).

The NUMAC instrumentation has certain design features which

contribute to its reliability. The replacement NUMAC LDMs are digital

instruments that use a microcomputer to monitor the ambient and

differential temperatures (also the MSL tunnel temperature timer) and

provide outputs and automatic self-testing and calibration. A

description of the major design features include: a) isolation of the

essential microcomputer by a serial data link from the front panel

display (and display microcomputer), b) a self-test system feature that

provides automatic testing of internal circuits and reports failures,

c) thermocouple failure detection, d) provisions to test the output

relays without the use of jumpers (reducing the threat of spurious

isolation), and e) two independent built-in instrument power supplies

(that automatically switchover to the other supply in the event of

failure). Also, several features have been included, among these, a) a

hardware ``watchdog'' timer to monitor against software cycling in

continuous loops, and b) software structured with the safety-related

essential tasks running at the highest priority in the system. These

capabilities increase the reliability of the collected data, reduce the

possibility of inadvertent isolation and plant shutdowns, reduce the

need for frequent calibrations, and reduce the likelihood of common

mode failures.

The NUMAC Leak Detection Monitors will maintain the same

environmental and electrical physical independence criteria

(qualifications) as the existing Leak Detection System components. The

LDMs, the associated thermocouple input units (TCIUs), and relay output

units (ROUs) will be mounted seismically such that qualification of

these components and the Control Room panels will be maintained. The

LDMs are qualified for the PNPP Control Room environment. The LDMs (one

per division) will be physically and electrically independent of each

other and do not share power supplies, thermocouple inputs, output

relays, microcomputer logic units, display units or enclosures and

mounting locations. A postulated gross failure of any one NUMAC LDM,

such as gross malfunction of the input unit, microcomputer logic unit

or the relay output unit, will not propagate to the other NUMAC LDM,

such as gross malfunction of the input unit, microcomputer logic unit

or the relay output unit, will not propagate to the other NUMAC LDM.

Thus, a failure within one NUMAC LDM will not prevent or disable the

function of the other NUMAC LDM. A failure within one NUMAC LDM may

cause the loss of one division of the isolation trip logic. However,

since the other redundant division (the MSLs have three other

divisions) will not be affected by this failure, the Leak Detection

System will still be able to perform its designed safety-related

function and provide the necessary system isolation. This is the same

as the current Leak Detection System design basis.

The possibility of a common mode failure of both NUMAC LDM

divisions is minimized by the design of the NUMAC hardware and

software, the verification and validation (V&V) of the software to

reduce the likelihood of errors, the testing of the software (to

discover and eliminate errors), and the design of and testing of the

hardware to demonstrate its resistance to EMI/RFI. The NUMAC instrument

design features, by effectively eliminating the potential for common

mode failures, maintain the Leak Detection System within its current

licensing basis. (A discussion of common mode failure protection is

presented in more detail in the answer to question two.) Therefore, the

design, isolation and separation criteria remain the same.

Additionally, as described within Chapter 7 of the Updated Safety

Analysis Report (USAR), diversity is provided to the ambient and

differential temperature monitoring trip functions for the various

systems by alternative leak detection methods (such as measuring steam

line flow or pressure) that provide backup in the event of the loss of

both divisions of the NUMAC Leak Detection Monitors. These alternative

leak detection methods are physically separate from those being

performed by the NUMAC LDM and constitute a diverse, redundant, safety

related backup capable of responding to a design basis line break for

the various systems. Therefore, a common mode failure of both LDM

divisions would not prevent any of the necessary system isolation from

occurring.

No changes are being made to the isolation logic of the Leak

Detection System. No accident initiators or precursors are affected by

the proposed changes to the Channel Functional Test surveillance

intervals for the various trip functions. One purpose of a Channel

Functional Test is to check the instrument setpoints. The NUMAC

instrument setpoints are set digitally and do not drift. An engineering

evaluation has established that the Channel Functional Test

surveillance interval can be extended from one to six months. The

potential for common mode failures has been accounted for in the design

and measures have been taken to lower the probability of this to an

acceptable level (see the answer to question two). Also, alternative

leak detection methods exist for this eventuality. Since the NUMAC Leak

Detection Monitoring equipment meets or exceeds the design and

licensing criteria specified for the Leak Detection System, the

proposed upgrade cannot increase the probability of occurrence of any

accident previously evaluated.

A portion of the Leak Detection System logic causes a closure of

the Main Steam Isolation Valves on a steam leak signal. This transient,

described in Chapter 15 of the USAR, may also occur due to a LDS

equipment malfunction. Since this modification replaces some of the

existing Riley temperature monitoring instrumentation with more

reliable instrumentation the probability of this transient is reduced

(no radiological consequences are associated with this event). The LDS

is also used to mitigate the consequences of a pipe break outside

primary containment by isolating the affected system connected to the

Reactor Coolant Pressure Boundary (RCPB). The replacement of the Riley

instrumentation with NUMAC LDMs will not change, degrade, or prevent

the Leak Detection System response to mitigate the radiological

consequences of an accident. Therefore, replacement of the Riley

temperature modules with NUMAC Leak Detection Monitors 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 single failure criterion requires that any single failure

within a safety-related system not prevent proper protective action of

the overall system when the system is required to function. The Leak

Detection System design is such that a failure of one division will not

prevent the system from performing its safety function. Common mode

failure protection provisions have been addressed in the NUMAC LDM

design.

The comprehensive General Electric software V&V and configuration

management control programs minimize, although they cannot entirely

eliminate, the likelihood of a common mode NUMAC instrument failure due

to software problems. The hardware (firmware) and software for the PNPP

NUMAC Leak Detection Monitors will undergo a formal software

verification and validation (V&V) process by General Electric, that is

to be completed by the end of the year, equivalent to the one reviewed

and approved by the NRC for the safety-related Wide Range Neutron

Monitor.

The NUMAC instruments are designed to minimize both their

susceptibility to, and generation of, electromagnetic and radio-

frequency interference (EMI/RFI) to prevent spurious operations and

allow their use in safety-related systems. As part of a broader plan by

GE to improve the testing has been performed by GE on the Leak

Detection Monitor configuration in order to both expand the overall

qualification region, and to obtain test data specific to this

application. This testing ensures the qualification of the Thermocouple

Input Unit (TCIU), a NUMAC circuit board which is unique to the LDM

application, and also extends the NUMAC EMI/RFI qualification region to

include both higher and lower frequencies than previously tested.

The NUMAC instrument design concept has undergone review by the

NRC, and the initial instruments of the NUMAC product line (the

Logarithmic Radiation Monitor and Wide Range Neutron Monitoring System)

have received NRC approval via Safety Evaluation of the associated GE

Licensing Topical Reports. The various types of NUMAC equipment in

operation at other nuclear power plants have components and software

modules which are similar to and in some instance identical to the

NUMAC LDMs. Therefore, based on the NRC reviewed and approved NUMAC

software and hardware control programs instituted by GE, the design

features to minimize software/hardware (or their interface) problems,

design features to minimize susceptibility to EMI/RFI, and testing to

demonstrate resistance to EMI/RFI, installation of NUMAC Leak Detection

Monitors at the PNPP does 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 replacement of the analog Riley temperature modules with the

microcomputer based NUMAC Leak Detection Monitors will not affect any

design conditions or impact the margins of safety for the various Leak

Detection System monitored parameters in the Technical Specification

Table 3.3.2-2 will not be changed or affected by this modification.

Only the CHANNEL FUNCTIONAL test interval is being extended.

The NUMAC Leak Detection Monitor design, with the attention paid

towards minimizing the potential for, and the effect of, software/

hardware and/or EMI/RFI related problem or common mode failures and

resulting operational experience has demonstrated that replacement of

the existing Riley temperature modules with NUMAC Leak Detection

Monitors would not result in a significant reduction in the margin of

safety.

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

this review, it appears that the three standards of 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.

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of amendment request: February 8, 1994, as supplemented March

25, 1994.

Description of amendment request: The amendment would revise the

WNP-2 Technical Specifications. Specifically, the amendment would

increase the stroke time, as specified in Table 3.6.3-1, for reactor

core isolation cooling (RCIC) valve RCIC-V-8, from 13 seconds to 26

seconds and the note (j) reference would be deleted from RCIC-V-8 and

RCIC-V-63. The note (j) indicates that the stroke time specified in the

Table reflects the requirement for containment isolation only.

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. This proposed action does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

RCIC-V-8 and V-63 are containment isolation valves and are normally

open. Failure of the valves to open or close cannot cause an accident.

The mitigating capability of RCIC-V-8 and V-63 is not changed in that

the valves will continue to be closed within the established time

limits. This ensures protection of the safety related equipment

necessary for continued compliance with the requirements of General

Design Criterion 4. In those accidents which involve a source term and

potential adverse dose release consequences, no credit is taken for the

closing of the valves; therefore the increase in the allowable time for

closing does not increase the consequences of those accidents.

2. This proposed action does not create the possibility of a new or

different kind of accident from any accident previously evaluated. The

requested Technical Specifications change does not represent a change

in modes of operation. It does not, in itself, require physical

modification to the plant, although it will be used to allow a gear

change in RCIC-MO-8. The new gears represent a standard configuration

for Limitorque motor operators and will require a routine design

change. The required Technical Specification change maintains the

licensing basis for the plant as discussed in response to question 1.

Hence, no new or different kind of accident is possible as a result of

implementing this change.

3. This proposed action does not involve a significant reduction in

a margin of safety. The increase in stroke time will increase the peak

temperature in the HELB profiles and thereby decrease the margin

available from the equipment qualification limits. However, sufficient

margin remains to assure the equipment operability is maintained and

there is no reduction in the margin of safety. Additionally, there is

no reduction in the margin of safety because increasing the stroke time

will not change the postulated radiological releases.

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: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Attorney for licensee: M.H. Philips, Jr., Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005-3502.

NRC Project Director: Theodore R. Quay.

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.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendments request: April 14, 1994.

Description of amendments request: The proposed change would

relocate the instrument response time tables 3.3.1-2, Reactor

Protection System (RPS) Response Times; 3.3.2-3, Isolation System

Instrumentation (ISI) Response Time; and 3.3.3-3, Emergency Core

Cooling System (ECCS) Response Times, from the Technical Specifications

to the Updated Final Safety Analysis Report. The RPS, ISI, and ECCS

instrument limiting conditions for operation (LCO) will be revised to

read that the instruments ``shall be operable'' without a reference to

a specific response time table in these LCOs. The references to the

response time tables will also be deleted from the Surveillance

Requirements.

Date of publication of individual notice in Federal Register: April

26, 1994 (59 FR 21785).

Expiration date of individual notice: May 26, 1994.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana

Date of application for amendment: January 14, 1994.

Brief description of amendment request: The proposed amendment

would revise various instrumentation technical specifications by

extending the allowable outage times (AOTs) of the instruments, and by

increasing their channel functional surveillance test intervals (STIs)

to quarterly. The amendment also revises certain technical

specification actions to address loss-of-function concerns associated

with the AOT and STI changes.

Date of individual notice in Federal Register: April 26, 1994(59 FR

21787).

Expiration date of individual notice: May 26, 1994.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana

Date of application for amendment: February 22, 1994.

Brief description of amendment request: The proposed amendment

would revise the technical specifications (TS) for the main steam-

positive leakage control system (MS-PLCS) and the penetration valve

leakage control system (PVLCS) to be consistent with the requirements

contained in NUREG-1434, ``Standard Technical Specifications, General

Electric Plants (BWR/6).''

Date of individual notice in Federal Register: March 10, 1994 (59

FR 11331).

Expiration date of individual notice: April 11, 1994.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana

Date of application for amendment: March 3, 1994.

Brief description of amendment request: The amendment would revise

the technical specifications in accordance with the guidance provided

by Generic Letter 93-08, ``Relocation of Technical Specification Tables

of Instrument Response Time Limits.'' Generic Letter 93-08 recommends

the removal and subsequent relocation of various technical

specification tables which denote instrument and system response time

limits.

Date of individual notice in Federal Register: March 16, 1994 (59

FR 12380).

Expiration date of individual notice: April 15, 1994.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

Street, NW., Washington, DC 20555, and at the local public document

rooms for the particular facilities involved.

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: October 26, 1993, as

supplemented March 28, 1994.

Brief description of amendments: The licensee is requesting a

revision to TS 5.3.1 for Palo Verde Nuclear Generating Station Units 1,

2, and 3 that will increase the maximum allowable fuel enrichment from

4.05 weight percent U-235 to 4.30 weight percent U-235. There was no

change requested to the current 52,000 MWD/MTU burnup. The licensee

provided a supplemental letter dated March 28, 1994, at the request of

the NRC to bring TS 5.3.1 into conformance with Generic Letter 90-02,

Supplement 1 and to clarify assumptions used in the Fuel Handling

Accident Analysis.

Date of issuance: April 19, 1994.

Effective date: April 19, 1994, to be implemented within 45 days of

issuance.

Amendment Nos.: 74, 60 and 46.

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 19, 1994 (59 FR

2860) The additional information contained in the supplemental letter

dated March 28, 1994, was clarifying in nature and thus within the

scope of the initial notice and did not affect the NRC staff's proposed

no significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated April 19, 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: September 1, 1992.

Brief description of amendments: The amendments revise Technical

Specification (TS) 3/4.4.3, ``Relief Valves,'' to improve the

reliability of the reactor coolant system's power-operated relief

valves (PORVs) and their associated block valves for overpressure

protection during normal operation and anticipated transients. The

amendments also revise TS 3/4.4.9, ``Pressure/Temperature Limits,'' to

improve the availability of the PORVs for low temperature overpressure

protection. Accompanying changes are also made to the associated TS

Bases. These revisions were made in response to Generic Letter 90-06,

``Resolution of Generic Issue 70, `Power-Operated Relief Valve and

Block Valve Reliability,' and Generic Issue 94, `Additional Low-

Temperature Overpressure Protection for Light-Water Reactors,' pursuant

to 10 CFR 50.54 (f).''

Date of issuance: April 20, 1994.

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

30 days.

Amendment Nos.: 188 and 165.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: September 30, 1992 (57

FR 45076).

The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated April 20, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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 application for amendment: February 4, 1994.

Brief description of amendment: The amendment revises TS

Surveillance Requirement 4.6.4.1 to delete the 12-hour channel check,

thereby eliminating the need for continuous operation of the hydrogen

monitors.

Date of issuance: April 26, 1994.

Effective date: April 26, 1994.

Amendment No. 47.

Facility Operating License No. NPF-63. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: March 2, 1994 (59 FR

10001) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 26, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

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

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 21, 1994, as supplemented

March 24, 1994.

Brief description of amendments: The amendments add a one-time

revision to Technical Specification (TS) 3/4.7.1.1 to permit continued

activities at all four units with main steam Code safety valve lift

setpoint tolerances of 3%. The duration of this amendment

is until May 9, 1994, at which time the tolerances will be reset to

1%. A statement has also been added to TS 4.7.1.1 for

Braidwood stating that the provisions of TS 4.0.4 are not applicable to

Braidwood, Unit 1, Cycle 5 until initial entry into Mode 2 from its

refueling outage.

Date of issuance: April 18, 1994.

Effective date: April 18, 1994.

Amendment Nos.: 61, 61, 49, and 49.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

Amendment revised the Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: Yes (59 FR 14685 dated March 29, 1994). The notice

provided an opportunity to submit comments on the Commission's proposed

no significant hazards consideration determination within 15 days. No

comments have been received. The notice also provided an opportunity to

request a hearing by April 29, 1994, but indicated that if the

Commission makes a final no significant hazards consideration

determination any such hearing would take place after issuance of the

amendment. The Commission's related evaluation of the amendment and

final no significant hazards consideration determination is contained

in a Safety Evaluation dated April 18, 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.

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

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

Date of application for amendment: September 29, 1993, as

supplemented by letter dated April 8, 1994.

Brief description of amendment: The amendment revises Technical

Specification Section 3.9.A.5 and Tables 3.9-1 and 4.10-2 to delete

controls for the 21, 22, and 23 Boron Monitor Tanks, which are no

longer in service.

Date of issuance: April 28, 1994.

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

the inlet and outlet lines of the 21, 22, and 23 Boron Monitor Tanks

have been cut and capped.

Amendment No.: 169.

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

Technical Specifications.

Date of initial notice in Federal Register: November 24, 1993 (58

FR 62154)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 28, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,

Michigan

Date of application for amendment: May 23, 1993.

Brief description of amendment: The amendment revises Technical

Specification (TS) 3.4.3.2.d and related Table 3.4.3.2-1 by changing

the allowable leakage for certain low pressure coolant injection (LPCI)

line pressure isolation valves and revises Table 3.6.3-1 to remove the

designation as containment isolation valves from the LPCI injection

reverse flow check and bypass valves. The related Bases are also

changed. Concurrently, the Commission granted an exemption from the

requirements of 10 CFR Part 50, Appendix J, III.C. for performing Type

C containment integrated leak rate tests of the containment isolation

valves in the low pressure coolant injection lines of the residual heat

removal system and to perform alternative testing.

Date of issuance: April 22, 1994.

Effective date: April 22, 1994, with full implementation within 45

days.

Amendment No.: 98.

Facility Operating License No. NPF-43. Amendment revises the

Technical Specifications

Date of initial notice in Federal Register: September 1, 1993 (58

FR 46227) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 22, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: February 24, 1993.

Brief description of amendment: The amendment corrected

typographical errors in the plant technical specifications (TSs). These

errors were introduced in the original ANO-1 TS, and in subsequent

amendments. These changes are administrative in nature and are intended

to improve the readability of the plant technical specifications

without changing the meaning or intent of any specifications.

Date of issuance: April 26, 1994.

Effective date: April 26, 1994.

Amendment No.: 171.

Facility Operating License No. DPR-51. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 22, 1993 (58

FR 67843)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 26, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: November 16, 1993, as supplemented by

letter dated April 5, 1994.

Brief description of amendment: The amendment revised the Technical

Specifications to provide acceptable conditions for operation when the

core operating limit supervisory system (COLSS) is out of service and

either or both control element assembly calculators (CEACs) are

operable.

Date of issuance: April 22, 1994.

Effective date: April 22, 1994.

Amendment No.: 93.

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

Technical Specifications.

Date of initial notice in Federal Register: January 5, 1994 (59 FR

620) The additional information contained in the supplemental letter

dated April 5, 1994, withdrew a portion of the original application and

thus, was 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 22, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: February 14, 1994.

Brief description of amendment: The amendment revised the Technical

Specifications in response to Generic Letter 93-08 issued by the NRC

and dated December 29, 1993, by relocating the reactor trip system and

engineered safety features actuation system response time limits to the

updated final safety analysis report.

Date of issuance: April 22, 1994.

Effective date: April 22, 1994.

Amendment No.: 94.

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

Technical Specifications.

Date of initial notice in Federal Register: March 16, 1994 (59 FR

12360) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 22, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

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: February 25, 1992.

Brief description of amendments: The amendments relate to your

application dated February 25, 1992, which requested a 40-year

operating license commencing from the date of issuance of the operating

license and, accordingly, would extend the operating license expiration

date for Turkey Point Units 3 and 4 to July 19, 2012 and April 10,

2013, respectively.

Date of issuance: April 20, 1994.

Effective date: April 20, 1994.

Amendment Nos. 162 and 156.

Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 15, 1992 (57 FR

13130) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 20, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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: December 28, 1993.

Brief description of amendments: These amendments include steam

generator overfill protection in the Technical Specifications in

response to Generic Letter 89-19, Request for Action Related to

Resolution of Unresolved Safety Issue A-47 ``Safety Implications of

Control Systems in LWR Nuclear Power Plants.''

Date of issuance: April 28, 1994.

Effective date: April 28, 1994.

Amendment Nos. 163 and 157.

Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: March 2, 1994 (59 FR

10007) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 28, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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

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

Date of application for amendment: March 14, 1994.

Brief description of amendment: The amendment changes the Millstone

Unit 2 Technical Specifications (TS) to provide a one-time extension of

the surveillance frequency from the required 18-month to the next

refueling outage but no later than September 30, 1994, of the power

operated valves in the service water system (TS 4.7.4.1.b) and in the

boron injection flowpath (TS 4.1.2.2.c). This extends the surveillance

for these valves approximately 5 months.

Date of issuance: April 22, 1994.

Effective date: April 22, 1994.

Amendment No.: 173.

Facility Operating License No. DPR-65. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 23, 1994 (59 FR

13751). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 22, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

Northeast Nuclear Energy Company, Docket Nos. 50-245, 50-336, and 50-

423, Millstone Nuclear Power Station, Units 1, 2 and 3, New London

County, Connecticut

Date of application for amendment: December 22, 1993.

Brief description of amendment: The amendments revise the Technical

Specifications (TS) as follows:

1. Change the title of the Nuclear Station Director to Senior Vice

President--Millstone Station.

2. Remove the requirement to provide a copy of Plant Operations

Review Committee (PORC) and Site Operations Review Committee (SORC)

meeting minutes to the Executive Vice President--Nuclear. The Senior

Vice President--Millstone Station replaces the Executive Vice

President--Nuclear for receipt of PORC and SORC meeting minutes.

3. Make editorial changes to the Millstone Unit No. 1 TS Index.

4. Correct a typographical error in Section 6.2.1.d of the

Millstone Unit No. 1 TS.

5. Correct a typographical error in Section 6.5.3.1.a of the

Millstone Unit No. 3 TS.

Date of issuance: April 26, 1994.

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

30 days.

Amendment Nos.: 74, 174, and 90.

Facility Operating License No. DPR-21, DPR-65, AND NPF-49.

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 16, 1994 (59

FR 7693) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 26, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendments: April 19, 1993, as supplemented

by letter dated April 18, 1994.

Brief description of amendments: These amendments extend

surveillance test interval and allowed outage times for the containment

isolation actuation instrumentation.

Date of issuance: April 26, 1994.

Effective date: April 26, 1994.

Amendment Nos. 69 and 32.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 23, 1993 (58 FR

34086) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 26, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendments: May 6, 1993, as supplemented by

letter dated April 18, 1994.

Brief description of amendments: These amendments extend

surveillance test interval and allowed outage times for selected

actuation instrumentation and makes editorial changes.

Date of issuance: April 26, 1994.

Effective date: April 26, 1994.

Amendment Nos. 70 and 33.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 23, 1993 (58 FR

34087) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 26, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Public Service Electric and Gas Company

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

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

Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: November 19, 1993.

Brief description of amendments: These amendments eliminate the

listing of specific position titles for the Plant Operations Review

Committee (PORC) composition in favor of allowing the Plant Manager to

appoint PORC members. This revision eliminates the need to change the

TS in the future whenever a position title is changed.

Date of issuance: April 26, 1994.

Effective date: April 26, 1994.

Amendments Nos.: 190 and 195.

Facility Operating License Nos. DPR-44 and DPR-56: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 5, 1994 (59 FR

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

in a Safety Evaluation dated April 26, 1994.

No significant hazards consideration comments received: No.

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.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: January 31, 1994.

Brief description of amendment: This amendment to the Appendix B

Technical Specifications (TSs), the Radiological Effluent TSs, revised

Section 3.5, and the associated Bases, to establish a threshold level

below which there will be no requirement to perform grab samples and

isotopic analyses of steam jet-air ejector (SJAE) effluent and revised

TS Table 3.10-1 to change the actions required when entering an SJAE

limiting condition for operation. Additionally, the amendment revised

the TSs to clarify instructions and make editorial corrections which

are administrative in nature.

Date of issuance: April 25, 1994.

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

30 days.

Amendment No.: 211.

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 2, 1994 (59 FR

10014).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 25, 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.

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

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

Date of application for amendment: March 12, 1993.

Brief description of amendment: The amendment revised the Technical

Specifications (TS) to incorporate the changes listed below:

(1) The frequency of high pressure water fire protection system

testing (specified in TS Section 4.12.A.1) was changed to accommodate

operation on a 24-month cycle.

(2) The frequency of fire pump diesel engine testing (specified in

TS Section 4.12.A.2) was changed to accommodate operation on a 24-month

cycle.

(3) The frequency of electrical tunnel, diesel generator building,

and containment fan cooler fire protection spray and/or sprinkler

system testing (specified in TS Section 4.12.B.1) was changed to

accommodate operation on a 24-month cycle.

(4) The frequency of fire barrier penetration seal inspection

(specified in TS Section 4.12.C.1) was changed to accommodate operation

on a 24- month cycle.

(5) The frequency of fire detection system testing (specified in TS

Section 4.12.D.1) was changed to accommodate operation on a 24-month

cycle.

(6) The frequency of fire hose station testing (specified in TS

Section 4.12.E.1) was changed to accommodate operation on a 24-month

cycle.

(7) The frequency of CO2 fire protection system testing

(specified in TS Section 4.12.G.1) was changed to accommodate operation

on a 24-month cycle. A new requirement was also added to exercise the

fire dampers on an annual basis.

These changes followed the guidance provided in Generic Letter 91-

04, ``Changes in Technical Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle,'' as applicable.

In addition, TS Section 4.12 was reformatted, in its entirety, and

several administrative changes were made to improve clarity.

Date of issuance: April 20, 1994.

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

30 days.

Amendment No.: 146.

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

Technical Specifications.

Date of initial notice in Federal Register: April 28, 1993 (58 FR

25862) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 20, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek

Generating Station, Salem County, New Jersey

Date of application for amendment: March 4, 1994.

Brief description of amendment: This amendment adds a new TS 3/

4.10.8, ``Inservice Leak and Hydrostatic Testing,'' to the Hope Creek

Generating Station TSs. The amendment also includes corresponding

changes to the TS Index, Table 1.2, ``OPERATIONAL CONDITIONS,'' and

provides Bases for TS 3/4.10.8. The added TS 3/4.10.8 permits the unit

to remain in OPERATIONAL CONDITION 4 with the average reactor coolant

temperature being increased above 200 deg.F, but not to exceed

212 deg.F, and certain OPERATIONAL CONDITION 3 Limiting Conditions for

Operation for secondary containment isolation, secondary containment

integrity and filtration, recirculation and ventilation system (FRVS)

operability being met.

Date of issuance: April 18, 1994.

Effective date: April 18, 1994.

Amendment No.: 69.

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 16, 1994 (59 FR

12384).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 18, 1994.

No significant hazards consideration comments received: None.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070.

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

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

Alabama

Date of application for amendments: December 23, 1992, as

supplemented August 12, 1993 and January 21, 1994 (TS 328).

Brief description of amendments: The amendments modify the

operability requirements for the low pressure coolant injection (LPCI)

mode of the residual heat removal (RHR) system while the reactor is

shut down. The amendments permit the RHR system to be considered

operable for LPCI when aligned for shutdown cooling if it can be

manually realigned and is not otherwise inoperable.

Date of issuance: April 19, 1994.

Effective date: April 19, 1994.

Amendment Nos.: 204, 223, and 177.

Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:

Date of initial notice in Federal Register: March 31, 1993 (58 FR

16873).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 19, 1994.

No significant hazards consideration comments received: None.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611.

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

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

Date of application for amendments: February 7, 1994 (TS 93-11).

Brief description of amendments: The amendments replace the wording

in Surveillance Requirement 4.7.9.i, ``Snubber Service Life Program,''

with that from the Westinghouse Electric Corporation Standard TS,

Revision 4a. In addition, the amendments delete the wording in SR

4.7.9.c, ``Snubber Visual Inspection Performance and Evaluation,'' that

is inconsistent with Generic Letter 90-09.

Date of issuance: April 18, 1994.

Effective date: April 18, 1994.

Amendment Nos.: 179--Unit 1 171--Unit 2.

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

revise the technical specifications.

Date of initial notice in Federal Register: February 7, 1994.

The Commission's related evaluation of the amendments are contained

in a Safety Evaluation dated April 18, 1994.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

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

Date of application for amendments: February 7, 1994 (TS 93-19).

Brief description of amendments: The amendments revise Technical

Specification 5.3.1 to allow the substitution of filler rods for fuel

rods in fuel assemblies by incorporating the guidance in Generic Letter

90-02, Supplement 1.

Date of issuance: April 18, 1994.

Effective date: April 18, 1994.

Amendment Nos.: 180--Unit 1 172--Unit 2.

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

revise the technical specifications.

Date of initial notice in Federal Register: March 16, 1994 (59 FR

12367) The Commission's related evaluation of the amendments are

contained in a Safety Evaluation dated April 18, 1994.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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 application for amendment: September 28, 1992.

Brief description of amendment: The amendment revised Technical

Specification (TS) 2.2, Limiting Safety System Settings, TS 3.3.1,

Reactor Protection System Instrumentation, and TS 3.3.2, Isolation

Actuation Instrumentation by removing the functions associated with the

main steam line radiation monitors.

Date of issuance: April 22, 1994.

Effective date: April 22, 1994.

Amendment No. 58.

Facility Operating License No. NPF-58. This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 6, 1993 (58 FR

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

in a Safety Evaluation dated April 22, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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 application for amendment: July 28, 1992, as supplemented

on February 17, 1993.

Brief description of amendment: The proposed amendment would delete

Technical Specification (TS) 3/4.9.9, ``Refueling Operations--

Containment Purge and Exhaust Isolation System,'' and its bases,

because of its redundancy to other TSs that address the operability

requirements of the containment purge and exhaust isolation system.

Also, the proposed amendment would revise TS 3/4.3.2, ``Safety System

Instrumentation--Safety Features Actuation System Instrumentation,''

and TS 3.4.9.4, ``Refueling Operations--Containment Penetrations,'' and

its bases. The effect of this proposed change would be to allow the

bypass of the safety features actuation system in Mode 6,

``Refueling,'' by the use of the containment purge and exhaust system

noble gas monitor in conjunction with manual closure of the containment

purge and exhaust isolation valves instead of automatic closure.

Date of issuance: April 15, 1994.

Effective date: April 15, 1994.

Amendment No. 186.

Facility Operating License No. NPF-3. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 6, 1993 (58 FR

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

in a Safety Evaluation dated April 15, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo Library,

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

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

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

Virginia

Date of application for amendments: October 8, 1993.

Brief description of amendments: The amendments revise the

technical specifications (TS) by deleting tables listing certain

components from the TS and relocating the lists to plant procedures in

accordance with the guidance provided in NRC Generic Letter 91-08,

``Removal of Component Lists from Technical Specifications.''

Date of issuance: April 22, 1994.

Effective date: April 22, 1994.

Amendment Nos.: 181 and 162.

Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1993 (58 FR

57860) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 22, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of application for amendment: July 29, 1993, as supplemented

by letters dated March 11 and 17, 1994.

Brief description of amendment: The amendment modifies the

Technical Specifications (TS) to reflect a new refueling mast.

Specifically, the amendment adds new values for protective features in

the TS to reflect the new refueling mast. Values for the old refueling

mast are retained in the TS.

Date of issuance: April 29, 1994.

Effective date: April 29, 1994.

Amendment No.: 121.

Facility Operating License No. NPF-21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 30, 1994 (59 FR

14900) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 29, 1994.

Public comments on proposed no significant hazards consideration

comments received: No.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

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

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

County, Wisconsin

Date of application for amendments: February 26, 1993.

Brief description of amendments: The amendments adding operating

conditions and limiting conditions for operation for the atmospheric

steam dump valves, the crossover steam dump system, the turbine stop

and governor valves, and the various turbine overspeed protection

features installed at the Point Beach Nuclear Plant. Additionally, the

amendments revised the surveillance requirements for the auxiliary

feedwater system, and added explanatory text to the bases for Sections

15.3.4 and 15.4.8.

Date of issuance: April 20, 1994.

Effective date: April 20, 1994.

Amendment Nos.: 147 and 151.

Facility Operating License Nos. DPR-24 and DPR-27. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 18, 1993 (58 FR

43939) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 20, 1994.

No significant hazards consideration comments received: No.

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

Sixteenth Street, Two Rivers, Wisconsin 54241.

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

Generating Station, Coffey County, Kansas

Date of amendment request: October 21, 1993, as supplemented by

letters dated March 14, 1994, and April 18, 1994.

Brief description of amendment: The amendment revises Technical

Specification Sections 6.5.1, Plant Safety Review Committee (PSRC) and

6.8, Procedures and Programs, in order to allow implementation of a

Qualified Reviewer Program for the review and approval of new

procedures and procedure changes. Technical Specification 6.5.1.6, PSRC

Responsibilities, has also been revised in accordance with Generic

Letter 93-07, ``Modification of Technical Specification Administrative

Control Requirements for Emergency and Security Plans,'' to delete

requirements for PSRC review of the Emergency Plan and Security Plan

and related implementing procedures.

Date of Issuance: April 28, 1994.

Effective date: April 28, 1994, to be implemented within 120 days

of issuance.

Amendment No.: Amendment No. 73.

Facility Operating License No. NPF-42. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 24, 1993 (58

FR 62159) The March 14, 1994, and April 18, 1994, supplemental letters

provided additional clarifying information and revised the

implementation period and did not change the initial no significant

hazards consideration. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated April 28, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room Locations: Emporia State University,

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

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

66621.

Dated at Rockville, Maryland, this 4th day of May 1994.

For the Nuclear Regulatory Commission.

Jack W. Roe,

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

Regulation.

[FR Doc. 94-11226 Filed 5-11-94; 8:45 am]

BILLING CODE 7590-01-P