Biweekly Notice

UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving

No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from June 13, 1994, through June 23, 1994. The

last biweekly notice was published on June 22, 1994 (59 FR 32226).

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 August 5, 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 John N. Hannon: 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.

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: June 2, 1994

Description of amendment request: The proposed amendment would

merge Toledo Edison Company into Cleveland Electric Illuminating

Company. As described in the application, the company formed from the

merger is intended to be renamed. Therefore, the licensee uses the

nomenclature ``NEWCO'' as a temporary name of the combined operating

company, and will provide the permanent name by supplemental letter.

The amendment would (1) replace the Toledo Edison Company and Cleveland

Electric Illuminating Company with ``NEWCO'' as a licensee, (2)

designate ``NEWCO'' as the owner of the Perry Nuclear Power Plant, Unit

1, and (3) make other administrative changes to the license as

indicated in the amendment application. Centerior Service Company would

be unaffected by the amendment and would remain a licensee.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes to the Operating License are administrative

and have no effect on any plant systems. All Limiting Conditions for

Operation, Limiting Safety Systems Settings and Safety Limits

specified in the Technical Specifications remain unchanged. This

change meets one of the examples of a change not likely to involve a

significant hazards consideration in that it is a purely

administrative changes (48 FR 14864).

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

or different kind of accident from any accident previously

evaluated.

The proposed changes to the Operating License are administrative

and have no effect on any plant systems. All Limiting Conditions for

Operation, Limiting Safety Systems Settings and Safety Limits

specified in the Technical Specifications remain unchanged. This

change meets one of the examples of a change not likely to involve a

significant hazards consideration in that it is a purely

administrative change (48 FR 14864).

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The proposed changes to the Operating License are administrative

and have no effect on any plant systems. All Limiting Conditions for

Operation, Limiting Safety Systems Settings and Safety Limits

specified in the Technical Specifications remain unchanged. This

change meets one of the examples of a change not likely to involve a

significant hazards consideration in that it is purely an

administrative change (48 FR 14864).

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

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

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

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

Will County, Illinois

Date of amendment request: May 20, 1994

Description of amendment request: The proposed amendment would

permit the licensee to use an alternate repair criteria (ARC),

designated as the F* criteria. Use of the F* criteria would

allow tubes with otherwise pluggable indications, to remain in service

as long as the indications are below the designated minimum distance of

the F* criteria. The F* criteria defines a length of 1.7

inches of undegraded expanded tube within the tubesheet as the minimum

distance acceptable for implementing this ARC. Below the F*

length, a circumferential tube defect can exist and the tube can remain

in service. The proposed amendment will change the plugging limit

definition and would exclude plugging steam generator tubes with

indications that satisfy the F* criteria. The F* criteria

maintains the structural integrity of the degraded tube as the primary

pressure boundary and allows the tube to remain in service for heat

transfer and core cooling.

This alternate repair criteria qualification is documented in

Babcock & Wilcox Nuclear Technologies (BWNT) Topical Report BAW-10196 P

Revision 1, ``W-D4 F* Qualification Report'', which is included as

part of the licensee's submittal.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The supporting qualification report for the subject criteria

demonstrates that the presence of the tubesheet will enhance the

tube integrity in the region of the tube-to-tubesheet roll

expansions by precluding tube deformation beyond its initial

expanded outside diameter. The resistance to a tube rupture is

strengthened by the presence of the tubesheet in that region. The

results of hardrolling of the tube into the tubesheet provides a

mechanical leak limiting seal between the tube and the tubesheet. A

tube rupture cannot occur because the contact between the tube and

the tubesheet does not permit sufficient movement of tube material.

The type of degradation for which the F* criteria has been

developed (cracking with a circumferential orientation) can

theoretically lead to a postulated tube rupture event provided that

the postulated through-wall circumferential crack exists near the

top of the tubesheet. An evaluation including analysis and testing

has been done to determine the resistive strength of the expanded

tubes within the tubesheet. This evaluation provides the basis for

the acceptance criteria for tube degradation subject to the F*

criteria.

The F* length of roll expansion is sufficient to preclude

tube pullout from tube degradation located below the F*

distance, regardless of the extent of the tube degradation. The

Technical Specification leakage rate requirements and accident

analysis assumptions remain unchanged in the unlikely event that

significant leakage from this region does occur. The tube rupture

and pullout is fully bounded by the existing steam generator tube

rupture analysis included in the UFSAR. The leakage testing of the

roll expanded tubes indicates that for tube expansion lengths

approximately equal to the F* distance, any postulated primary

to secondary leakage from F* tubes would be insignificant. The

proposed alternate repair criteria does not adversely impact any

other previously evaluated design basis accident.

The leakage from an F* tube would be limited by the tube-

to-tubesheet interface since this leak would occur below the

secondary face of the tubesheet. Qualification testing and previous

experience indicate that normal and faulted leakage is well below

Technical Specification and administrative limits creating no

increase in the consequences associated with tube rupture type

leakages. The UFSAR analyzed accident scenarios are still bounding

since the normal and faulted leak rates are well within the normal

operating limit of 150 gallons per day. This conclusion is

consistent with previous F* programs approved and used at other

operating plants.

All of the design and operating characteristics of the steam

generator and connected systems are preserved since the F*

criteria utilizes the ``as rolled'' tube configuration that exists

as part of the original steam generator design. The F* joint

has been analyzed and tested for design, operating, and faulted

condition loadings in accordance with Regulatory Guide 1.121 safety

factors. The potential for tube rupture is not increased from the

original submittal as demonstrated in the qualification analyses and

testing completed in the BWNT report.

Therefore, this change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

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

or different type of accident from any accident previously

evaluated.

Implementation of the proposed F* criteria does not

introduce any significant changes to the plant design basis. Use of

the criteria does not provide a mechanism to initiate an accident

outside of the region of the expanded portion of the tube. In the

unlikely event the failed tube severed completely at a point below

the F* region, the remaining F* joint would retain

engagement in the tubesheet due to its length of expanded contact

within the tubesheet bore. This engagement length would prevent any

interaction of the severed tube with neighboring tubes. Any

hypothetical accident as a result of any tube degradation in the

expanded region of the tube would be bounded by the existing tube

rupture accident analysis. Tube bundle structural integrity will be

maintained. Tube bundle leak tightness will be maintained such that

any postulated accident leakage from F* tubes will be

negligible with regard to offsite doses.

Therefore, there is not a potential for creating the possibility

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

evaluated.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

The use of the F* criteria has been demonstrated to

maintain the integrity of the tube bundle commensurate with the

requirements of Reg Guide 1.121 and the primary to secondary

pressure boundary under normal and postulated accident conditions.

Acceptable tube degradation for the F* criteria is any

degradation indication in the tubesheet region, more than the

F* distance from the secondary face of the tubesheet or the top

of the last hardroll contact point which ever is further into the

tubesheet. The safety factors used in the verification of the

strength of the degraded tube are consistent with the safety factors

in the ASME Boiler and Pressure Vessel Code and Reg Guide 1.121 used

in steam generator design. The F* distance has been verified by

various testing to be greater than the length of the roll expanded

tube-to-tubesheet interface required to preclude both tube pullout

and significant leakage during normal and postulated accident

conditions. The protective boundaries of the steam generator

continue to be maintained with the use of the F* criteria. A

tube with the indication of degration previously requiring removal

from service can be kept in service through the F* criteria.

Since the joint is constrained within the tubesheet bore, there is

no additional risk associated with the previously analyzed tube

rupture event. The leak testing acceptance criteria are based on the

primary to secondary leakage limit in the Technical Specifications

and the leakage assumptions used in the UFSAR accident analyses.

Implementation of the alternate repair criteria will decrease

the number of tubes which must be taken out of service with tube

plugs or repaired by sleeves. Both plugs and sleeves reduce the RCS

flow margin; thus, implementation of the F* criteria will

maintain the margin of flow that would otherwise be reduced in the

event of increased plugging or sleeving.

Based on the above, it is concluded that the proposed change

does not result in a significant reduction in margin with respect to

plant safety as defined in the UFSAR or the Technical Specification

Bases.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: For Byron, the Byron Public

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

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

Street, Wilmington, Illinois 60481

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: October 22, 1993

Description of amendment request: The proposed amendment would

modify the Reactor Trip System (RTS) and Engineered Safety Feature

(ESF) instrumentation surveillance requirements to incorporate the

applicable changes specified in NRC-approved WCAP-10271 and related

supplements. Four specific changes were approved by the Nuclear

Regulatory Commission for the RTS and Engineered Safety Feature

Actuation System analog channels. These changes are limited to the

specific Reactor Protection System (RPS) channels evaluated in the WCAP

(including all supplements) and are subject to the conditions specified

by the NRC.

1. The surveillance or test frequency may be changed from monthly

to quarterly.

2. The time allowed for a channel to be inoperable or out of

service in an untripped condition may be changed from 1 hour to 6

hours.

3. The time a channel in a functional group may be bypassed to

perform testing may be increased from 2 to 4 hours. This bypass time

applies to either an inoperable channel when testing is done in the

tripped mode or to the channel in test when testing is done in the

bypass mode. The Allowed Outage Time for maintenance of a channel is 12

hours.

4. Routine channel testing may be performed in the bypassed

condition instead of the tripped condition.

In addition, a number of editorial changes are made to improve

clarity, and two-loop operating requirements are proposed to be

deleted. Also, the surveillance test interval for RTS interlocks is

proposed to be changed from monthly to once-per-refueling (about 18

months). Although not part of WCAP-10271, this was previously approved

generically by NRC.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The determination that the results of the proposed changes are

within all acceptable criteria was established in the SER(s) [Safety

Evaluation Report(s)] prepared for WCAP-10271, WCAP-10271 Supplement

1, WCAP-10271 Supplement 2 and WCAP-10271 Supplement 2, Revision 1

issued by letters dated February 21, 1985, February 22, 1989 and

April 30, 1990. Implementation of the proposed changes is expected

to result in an acceptable increase in total Reactor Protection

System yearly unavailability. This increase, which is primarily due

to less frequent surveillance, results in an increase of similar

magnitude in the probability of an Anticipated Transient Without

Scram (ATWS) and in the probability of core melt resulting from an

ATWS and also results in a small increase in Core Damage Frequency

(CDF) due to Engineered Safety Features Actuation System

unavailability.

Implementation of the proposed changes is expected to result in

a significant reduction in the probability of core melt from

inadvertent reactor trips. This is a result of a reduction in the

number of inadvertent reactor trips (0.5 fewer inadvertent reactor

trips per unit per year) occurring during testing of RPS

instrumentation. This reduction is primarily attributable to testing

in bypass and less frequent surveillance.

The reduction in [***] core melt frequency is sufficiently large

to counter the increase in ATWS core melt probability resulting in

an overall reduction in total core melt probability.

The values determined by the WOG [Westinghouse Owners Group] and

presented in the WCAP for the increase in CDF were verified by

Brookhaven National Laboratory (BNL) as part of an audit and

sensitivity analyses for the NRC Staff. Based on the small value of

the increase compared to the range of uncertainty in the CDF, the

increase is considered acceptable.

Changes to Surveillance Test Frequencies for the Reactor Trip

System Interlocks do not represent a significant reduction in

testing. The currently specified test interval for interlock

channels allows the surveillance requirement to be satisfied by

verifying that the permissive logic is in its required state using

the annunciator status light. The surveillance, as currently

required, only verifies the status of the permissive logic and does

not address verification of channel setpoint or operability. The

setpoint verification and channel operability are verified after a

refueling shutdown. The definition of the channel check includes

comparison of the channel status with other channels for the same

parameter. The requirement to routinely verify permissive status is

a different consideration than the availability of trip or actuation

channels which are required to change state on the occurrence of an

event and for which the function availability is more dependent on

the surveillance interval. The change in surveillance requirement to

at least once every 18 months does not therefore represent a

significant change in channel surveillance and does not involve a

significant increase in unavailability of the Reactor Protection

System.

The proposed changes do not result in an increase in the

severity or consequences of an accident previously evaluated.

Implementation of the proposed changes affects the probability

of failure of the RPS but does not alter the manner in which

protection is afforded nor the manner in which limiting criteria are

established.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes do not result in a change in the manner in

which the Reactor Protection System provides plant protection. No

change is being made which alters the functioning of the Reactor

Protection System (other than in a test mode). Rather the likelihood

or probability of the Reactor Protection System functioning properly

is affected as described above. Therefore,the proposed changes do

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

nor involve a reduction in a margin of safety as defined in the

Safety Analysis Report.

The proposed changes do not involve hardware changes except

those necessary to implement testing in bypass. Some existing

instrumentation is designed to be tested in bypass and current

technical specifications allow testing in bypass. Testing in bypass

is also recognized by IEEE [Institute of Electrical and Electronics

Engineers] Standards. Therefore, testing in bypass has been

previously approved and implementation of the proposed change for

testing in bypass does not create the possibility of a new or

different kind of accident from any previously evaluated.

Furthermore, since the other proposed changes do not alter the

functioning of the RPS, the possibility of a new or different kind

of accident from any previously evaluated has not been created.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes do not alter the manner in which safety

limits, limiting safety system setpoints or limiting conditions for

operation are determined. The impact of reduced testing other than

as addressed above is to allow a longer time interval over which

instrument uncertainties (e.g., drift) may act. Experience has shown

that the initial uncertainty assumptions are valid for reduced

testing.

Implementation of the proposed changes is expected to result in

an overall improvement in safety by:

a. Less frequent testing will result in fewer inadvertent

reactor trips and fewer actuations of Engineered Safety Feature

Actuation System components.

b. Improvements in the effectiveness of the operating staff in

monitoring and controlling plant operation. This is due to less

frequent distraction of the operator and shift supervisor to attend

to instrumentation testing.

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: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,

Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Walter R. Butler

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: April 19, 1994

Description of amendment request: The proposed amendment would

modify Specifications 3.4.9.3 and 3.4.11 to incorporate power operated

relief valve (PORV) Technical Specification (TS) changes in accordance

with the guidance in Generic Letter 90-06 as implemented in NUREG-1431

Improved Standard Technical Specifications (ISTS), with some exceptions

and modifications to reflect plant specific design features. Certain

other TS sections would also be modified to address related TSs.

The proposed changes involve the details of (a) limiting conditions

of operation, and (b) surveillance testing for equipment needed to

protect the reactor vessel from overpressure conditions. This equipment

includes PORVs and their associated block valves, charging pumps,

reactor coolant system (RCS) vent, accumulators, and the overpressure

protection system. Numerous administrative changes are also proposed,

such as renumbering sections, spelling out mathematical symbols,

changes in nomenclature for consistency, and relocating sentences and

paragraphs.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes consolidate the power operated relief valve

requirements into Specifications 3.4.9.3 and 3.4.11 which generally

adopt the new Improved Standard Technical Specifications of NUREG-

1431 to address the concerns identified in Generic Letter 90-06

except for those changes required to reflect plant specific design

features. These changes are proposed to enhance safety and improve

the reliability of the PORVs and block valves. Since the proposed

changes augment or preserve the requirements contained in the

current technical specifications, we have concluded that these

changes do not involve a significant increase in the probability or

consequences of an accident previously evaluated in the UFSAR

[Updated Final Safety Analysis Report].

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes do not involve any physical changes to the

PORVs or their setpoints. These changes do not delete any function

previously provided by the PORVs nor has the probability of

inadvertent opening been increased. Accordingly, no new failure

modes have been defined for any plant system or component important

to safety nor has any new limiting single failure been identified as

a result of these changes. Therefore, these changes will not create

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

accident previously evaluated in the UFSAR.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes have been incorporated to enhance safety

and improve the reliability of the PORVs and block valves to ensure

their availability when called upon to perform their function. These

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

or involve a change to equipment or features which affect the

operational characteristics of the facility. Therefore, operation of

the facility in accordance with the proposed amendment would not

involve a significant reduction in a margin of safety.

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

that a significant effort has been made by the NRC and by industry over

the last several years to improve and tighten the TS requirements for

overpressure protection systems. These efforts are documented in

Generic Letter 90-06 and NUREG-1431. The changes proposed by the

licensee appear to result in TSs that are significantly more

comprehensive and restrictive than those now existing for Beaver Valley

Units 1 and 2, and, therefore, should help to reduce the probability of

an accident. The staff disagrees with the licensee's claim that the

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

(consideration number 3 above). However, the staff believes that the

proposed TS changes require more restrictive operation (such as more

careful control of the number of charging pumps which can inject into

the RCS) and do not involve a significant reduction in a margin of

safety. Based on the NRC staff's 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: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,

Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Walter R. Butler

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: June 9, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification section 4.8.1.1.2 to replace the current

qualitative examination of new diesel generator fuel oil for water/

sediment and particulate contamination with a quantitative examination

for the same properties.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated

since the diesel generator availability and reliability is not being

changed. The quantitative acceptance criteria for new fuel oil is

not being changed. Diesel generator performance will therefore not

be changed due to the proposed revision to SR [surveillance

requirement] 4.8.1[.1].2.d.1.d. The diesel generator will continue

to provide sufficient electrical power to ESF [engineered safety

feature] systems. The ESF systems will continue to function, as

assumed in the safety analyses, to ensure that the fuel, reactor

coolant system, and containment

design limits are not exceeded.

Therefore, this changes will not increase the probability or

consequences of an accident previously evaluated due to the

continued availability and reliability of the emergency diesel power

source.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed change does not alter the method of operating the

plant. This change will continue to ensure that the addition of new

fuel oil complies with accepted standards regarding fuel oil

quality. Since design requirements continue to be met and the

integrity of the reactor coolant system pressure boundary is not

challenged, no new failure mode has been created. As a result, an

accident which is different than any already evaluated in the

Updated Final Safety Analysis Report will not be created due to this

change.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The margin of safety is not reduced because the emergency diesel

generators will continue to provide sufficient capacity, capability,

redundancy, and reliability to ensure availability of necessary

power to ESF systems. The ESF systems will continue to function, as

assumed in the safety analyses, to ensure that the fuel, reactor

coolant systems, and containment design limits are not exceeded. The

replacement of the clear and bright qualitative examination with the

proposed quantitative test to determine the actual water/sediment

and particulates will ensure that new fuel oil meets the required

limits for these properties prior to addition to the storage tank,

therefore assuring that the quality of the stored fuel is unaffected

by the addition of new fuel.

Therefore, this proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,

Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Walter R. Butler

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley

Power Station, Unit No. 2, Shippingport, Pennsylvania

Date of amendment request: February 16, 1994

Description of amendment request: The proposed amendment would

delete the Appendix B Section 4.2.2 requirement to perform infrared

aerial photography every other year.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change will delete from Facility Operating License

No. NPF-73 the Appendix B Section 4.2.2 requirement to perform

infrared aerial photography every other year. The acceptance limit

which forms the licensing basis for this technical specification is

related to environmental impact and has no impact on the margin of

safety, accident analysis, or other design basis impacting the

margin of safety. No increase in adverse environmental impact has

been identified over that previously identified in the Final

Environmental Statement - Operating License Stage, environmental

impact appraisals, or in any decisions of the Atomic Safety and

Licensing Board. The Final Environmental Statement concluded, based

on a model of combined drift from Units 1 and 2, that no adverse

impacts to sensitive species of natural vegetation or to sensitive

species of crops were expected. The staff also examined infrared

aerial photographs taken from 1975 through 1983 and found no injury

to vegetation from cooling tower drift in the vicinity of Unit 1.

Continued terrestrial monitoring was performed for Beaver Valley

Unit 2 by infrared aerial photography in 1986, 1988, 1990, and 1992.

The results as provided in the Annual Environmental Reports Non-

Radiological concluded, ``Based on interpretation of the infrared

photographs and field verification, there is no evidence to suggest

that the BVPS [Beaver Valley Power Station] cooling towers are

causing vegetation stress.''

Based on the compilation of the infrared aerial photography

performed for both BV-1 and BV-2, deletion of this terrestrial

monitoring requirement will have no impact on the environment or the

operation of the plant. Therefore, the proposed change will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The Appendix B Section 4.2.2 requirement to perform infrared

aerial photography reflects a commitment described in Final

Environmental Statement Section 5.14.1. Therein it is stated that

the preoperational monitoring studies for BV-2 are based primarily

on the BV-1 operational monitoring programs. ``Results of these

studies have shown that there were no BV-1 operational impacts on

flora, thus, the only terrestrial monitoring planned for BV-2 is

continued infrared aerial photography every other year. The

photographs will be compared with preoperational photographs of the

BV-2 area, nd any signs of injury as a result of salt drift and

other sources will be checked. The details of this terrestrial

monitoring program will be specified in the Environmental Protection

Plan that will be included in Appendix B of the operating license.''

The subject of this concern is the impact of salt and water drift on

area vegetation including sensitive agricultural crops. From the

standpoint of soil salinization (the effects of the accumulation of

salts in the soil), described in the Environmental Report-Operating

License Stage Section 5.3.3, no appreciable impact resulting from

operation of the natural draft cooling towers is anticipated. This

is based on the average rate of precipitation of 36.2 inches

annually which greatly reduces the potential for accumulation of

salt in the soil. The terrestrial monitoring program has been

performed in accordance with the Environmental Protection Plan and

has provided additional verification that operation of both cooling

towers has not produced any evidence of vegetation stress. The

proposed change does not introduce any new mode of plant operation

or require any physical modification to the plant, therefore, this

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

accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

Infrared aerial photography surveillance does not affect safety

systems and/or systems important to safety. Terrestrial monitoring

is not used in any accident analysis and does not provide a basis

for evaluating the radiological consequences of an accident.

Deleting the requirement to perform infrared aerial photography will

not result in any environmental impact from operation of the cooling

tower and will not affect the operation of the cooling tower. The

operating history of both the BV-1 and the BV-2 cooling towers has

demonstrated that there is no evidence of vegetation stress in

accordance with the results obtained from the infrared aerial

photography and other associated methods of environmental

monitoring. Therefore, based on the above, the proposed change will

not involve a significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,

Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Walter R. Butler

Florida Power and Light Company, et al., Docket No. 50-335, St.

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

Date of amendment request: May 23, 1994

Description of amendment request: The amendment will revise

Technical Specification (TS) 3.5.2 for Emergency Core Cooling Systems

(ECCS) by removing the option that allows High Pressure Safety

Injection (HPSI) Pump 1C to be used as an alternative to the preferred

pump for subsystem operability. HPSI pump 1C is an installed spare

which is not required to be maintained in an operable status, and this

change is being requested to upgrade the ECCS operability requirements

consistent with actual plant operating needs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed license amendment will remove the option of using

High Pressure Safety Injection Pump 1C (HPSI-1C) to satisfy, in

part, the Emergency Core Cooling System (ECCS)

operabilityrequirements specified in Limiting Condition for

Operation (LCO) 3.5.2. HPSI-1C is an installed spare pump that is

not required to be operable unless it is being used in place of the

preferred B-train ECCS high pressure pump. The required functional

response of the ECCS or the required availability of the minimum

equipment necessary to accomplish the ECCS safety function will not

be changed by removing the spare pump option from the Technical

Specifications.

The calculated cooling performance of the St. Lucie Unit 1 ECCS

during postulated accidents conforms to the criteria set forth in 10

CFR 50.46 and the ability to achieve this required performance,

including considerations of single-failure criteria, is independent

from optional use of HPSI-1C. Therefore, operation of the facility

in accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment will not change the physical plant or the

modes of operation defined in the facility license. Eliminating the

option for the licensee to utilize HPSI-1C in place of the preferred

B-train ECCS high pressure pump does not involve the addition of new

or different types of equipment to the previously analyzed ECCS.

Equipment important to safety will continue to perform their safety

functions as previously analyzed and will not be affected by this

proposed amendment. Therefore, operation of the facility in

accordance with the proposed amendment would not create the

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

previously evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

Removing the option to employ the installed spare HPSI pump 1C

in lieu of the preferred B-train pump to determine ECCS operability

only removes an operational flexibility that has rarely been used by

the licensee. St. Lucie Unit 1 accident analyses do not take credit

for an installed spare pump, the minimum complement of safety

injection equipment required for safe operation of the facility and

that is required by LCO 3.5.2 is not changed, and the results of

plant accident and transient analyses are not influenced by this

proposed amendment. The proposed change does not alter the bases for

any Technical Specification related to the establishment of, or

maintenance of, a nuclear safety margin. Therefore, operation of the

facility in accordance with the proposed amendment would not involve

a significant reduction in a margin of safety.

Based on the discussion presented above and on the supporting

Evaluation of Proposed TS Changes, FPL has concluded that this

proposed license amendment involves no significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

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

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

NRC Project Director: Herbert N. Berkow

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: May 23, 1994

Description of amendment request: The proposed amendments will

relocate the seismic monitoring instrumentation Limiting Conditions for

Operation, Surveillance Requirements, and the associated tables

contained in Technical Specifications 3.3.3.3, 4.3.3.3.1 and 4.3.3.3.2

to the Updated Final Safety Analysis Report. The basis for this request

is consistent with NUREG-1432, ``Standard Technical Specifications,

Combustion Engineering Plants'' and with the ``Final Policy Statement

on Technical Specifications Improvements for Nuclear Power Reactors,

``published in the Federal Register (58 FR 39132) dated July 22, 1993.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes are administrative in nature in that the

specifications for operation and surveillance of the Seismic

Monitoring Instrumentation system will be relocated from Appendix A

of the facility operating license to the Updated Final Safety

Analysis Report for St. Lucie Unit 1 and Unit 2. Changes to the

system will be controlled by 10 CFR 50.59 and the safety analysis

report is required to be updated pursuant to 10 CFR 50.71(e).

Relocation of these requirements to the UFSAR is consistent with the

NRC ``Final Policy Statement on Technical Specifications

Improvements for Nuclear Power Reactors'' published in the Federal

Register (58 FR 39132) dated July 22, 1993.

Seismic monitoring instrumentation is not an accident initiator

nor a part of the success path(s) which function to mitigate

accidents evaluated in the plant safety analyses. The proposed

technical specification change does not involve any change to the

configuration or method of operation of any plant equipment that is

used to mitigate the consequences of an accident, nor do the changes

alter any assumptions or conditions in any of the plant accident

analyses. Therefore, operation of the facility in accordance with

the proposed amendment would not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment to relocate the existing Technical

Specification requirements for Seismic Monitoring Instrumentation to

the Updated Final Safety Analysis Report will not change the

physical plant or the modes of plant operation defined in the

Facility License. The change does not involve the addition or

modification of equipment nor does it alter the design or operation

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

with the proposed amendment would not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

The proposed changes are administrative in nature in that

operating and surveillance requirements for the Seismic Monitoring

Instrumentation system will be relocated from Appendix A of the

facility license to the Updated Final Safety Analysis Report for St.

Lucie Unit 1 and Unit 2. Seismic monitoring instruments are not used

to actuate safety-related equipment, provide interlocks, or

otherwise perform plant control functions. The instruments are used

to record the magnitude of a seismic event, should it occur.

Conditions evaluated in plant accident and transient analyses do not

involve seismic instruments. The proposed changes do not alter the

basis for any technical specification that is related to the

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

Therefore, operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

Based on the above discussion and the supporting Evaluation of

Technical Specification changes, FPL has determined that the

proposed license amendment involves no significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

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

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

NRC Project Director: Herbert N. Berkow

Florida Power and Light Company, et al., Docket No. 50-389, St.

Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: May 23, 1994

Description of amendment request: The proposed amendment revises

Technical Specifications Section 3/4.7.1.1, Turbine Cycle, Safety

Valves, to delete a specific reference to the 1974 edition of the ASME

Code and refer to testing in accordance with Technical Specification

4.0.5, the In-Service Inspection and In-Service Testing Specification.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated because the main steam code safety valves will continue to

be tested in accordance with current NRC requirements as implemented

through 10 CFR 50.55a. The NRC specifies the ASME code requirements

for a facility through revisions to 10 CFR 50.55a and through the

review and approval of the plant specific in-service testing plan

for pumps and valves at the beginning of each in-service inspection

interval.

The probability or consequences of an accident are not increased

because testing of the main steam safety valves is in accordance

with the appropriate NRC requirements.

(2) Use of the modified specification would not create the

possibility of a new or different kind of accident from any

previously evaluated.

The use of this modified specification can not create the

possibility of a new or different kind of accident from any

previously evaluated since there is no physical change to the

facility or the set points for the main steam safety valves. The

valves will be tested in accordance with current requirements. No

new failure mode is introduced due to the change because no plant

change is being made and main steam safety valve test methods are

consistent with the endorsed edition of the ASME Code.

(3) Use of the modified specification would not involve a

significant reduction in a margin of safety.

The existing technical specification references an outdated

version of the ASME Code. This change corrects the reference to

Specification 4.0.5 which ensures that in-service testing of ASME

Code Class 1, 2, and 3 pumps and valves will be performed in

accordance with a periodically updated version of Section XI of the

ASME Boiler and Pressure Vessel Code and Addenda as required by 10

CFR 50.55a.

Safety valve setpoints or tolerances are not changed by this

proposal. Therefore, the modified specification corrects the ASME

Code reference and does not involve a significant reduction in a

margin of safety.

Based on the above, we have determined that the proposed

amendment does not (1) involve a significant increase in the

probability or consequences of an accident previously evaluated, (2)

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

any previously evaluated, or (3) involve a significant reduction in

a margin of safety; and therefore does not involve a significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

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

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

NRC Project Director: Herbert N. Berkow

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

Linn County, Iowa

Date of amendment request: March 27, 1992, as supplemented on

January 6, and May 27, 1994.

Description of amendment request: The proposed amendment would

revise the limiting conditions for operation and surveillance

requirements for primary containment integrity, secondary containment

integrity and other systems and equipment of Technical Specifications

Section 3.7.

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 will not involve a significant

increase in the probability or consequences of an accident

previously evaluated because the requested revisions do not affect

the FSAR safety analyses involving these system.

Definitions

The revisions to Definition 15, ``Primary Containment

Integrity'' and Definition 16, ``Secondary Containment Integrity''

agree with the corresponding definitions of the STS. These changes

are administrative in nature in that they only clarify the

requirements for containment integrity and the appropriate means of

isolating penetrations. These changes do not affect the operation or

function of the containment isolation systems and, therefore, do not

result in a significant increase in the probability or consequences

of an accident previously evaluated.

Primary Containment Integrity

The revision to TS section 3.7.A, ``Primary Containment

Integrity'', only adds a specific requirement to restore primary

containment integrity within 1 hour or commence a plant shutdown.

These actions are consistent with the actions specified in STS for

primary containment integrity. No changes to the primary containment

boundary or the requirements for primary containment integrity have

been proposed. Therefore, this change does not result in a

significant increase in the probability or consequences of an

accident previously evaluated.

Primary Containment Power Operated Isolation Valves

The revisions to TS section 3.7.B, ``Primary Containment Power

Operated Isolation Valves'' are editorial in nature in that the

wording has only been changed to be consistent with the STS

requirements for primary containment isolation valves. These changes

do not affect the function of the valves, the requirements to

isolate a penetration with an inoperable containment isolation valve

or the actual methods of isolation. Penetrations are still required

to be isolated within 4 hours in a manner that cannot be adversely

affected by a single active failure. Therefore, these changes do not

result in a significant increase in the probability or consequences

of an accident previously evaluated.

Drywell Average Air Temperature

The addition of limits, actions, and surveillance requirements

for drywell average air temperature are intended to ensure that the

initial assumptions in the DAEC Primary Containment Response

Analysis to a DBA remain valid. The temperature limit (135 deg.F)

corresponds to the initial drywell average temperature assumed for

this analysis in the UFSAR. The specified limits, actions and

surveillance requirements are consistent with STS. The addition of

this limit to the TS will not affect the actual operation or

function of any equipment but will ensure that the containment

analysis remains valid. Therefore, the addition of this limit will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

Pressure Suppression Chamber-Reactor Building Vacuum Breakers

The changes to TS section 3/4.7.D only provide additional detail

and operability requirements for the pressure suppression chamber-

reactor building vacuum breakers. These additional details are

consistent with the requirements of STS. Specifying separate

operability requirements for vacuum breakers inoperable for opening

(but known to be closed), or open better reflects the dual functions

of these valves (vacuum relief and containment isolation). The

additional surveillance requirement will better ensure that the

containment isolation function of these valves is maintained. The

rewording of existing surveillances only clarifies current

requirements. These changes do not affect the actual function,

setpoints, or number of valves required to be operable and therefore

do not result in a significant increase in the probability or

consequences of an accident previously evaluated.

Drywell-Pressure Suppression Chamber Vacuum Breakers

The changes to TS section 3/4.7.E, only provide additional

detail and operability requirements for the drywell-pressure

suppression chamber vacuum breakers. These additional details are

consistent with the requirements of STS. Specifying separate

operability requirements for vacuum breakers inoperable for opening

(but known to be closed) or open better reflects the dual functions

of these valves. The additional requirement to verify that each

vacuum breaker is closed at least once per week will better ensure

that the isolation boundary between the drywell and torus is

maintained. The elimination of the requirement to exercise all

operable drywell-pressure suppression chamber vacuum breakers upon

determination that a vacuum breaker is inoperable for opening will

not affect the reliability of these vacuum breakers. The only valid

reason to exercise the operable vacuum breakers is if a common mode

failure is suspected. We have reviewed the maintenance history of

these valves and have not identified any instance of common mode

failures. Conditional testing of these changes will not result in a

significant increase in the probability or consequences of an

accident previously evaluated.

Main Steam Isolation Valve Leakage Control System (MSIV-LCS)

The change to TS Section 3/4.7.F, ``MSIV-LCS'', deletes the

unnecessary and potentially non-conservative conditional

surveillance testing of the redundant MSIV-LCS subsystems. Although

the proposed change will reduce the amount of testing of the MSIV-

LCS, reliability of these systems would not be decreased and the

necessary assurance that the alternate systems/subsystems/components

will operate when needed is provided by the ASME Section XI IST

Program.

The possibility of human error will decrease with reduced

testing. Human error such as a misalignment of valves after the

system is returned to its normal configuration following testing and

the misdirection of the operators attention from monitoring and

directing plant operations is less likely to occur if this testing

is eliminated. Additionally, reducing the scope and frequency of

surveillance testing will decrease the probability of equipment

failure (due to testing) which could require plant shutdown.

Therefore, this change will not increase the probability of

occurrence or consequences of an accident previously evaluated.

Suppression Pool Level and Temperature

The changes to TS section 3/4.7.F, ``Suppression Pool Level and

Temperature'', are intended to clarify these requirements and make

them more consistent with STS. The revision to the applicability

statement which deletes the requirement for suppression pool level

and temperature to be within the specified limits during work which

has the potential to drain the vessel is in accordance with STS.

Suppression pool level and temperature limits ensure that the

suppression pool has the capability of acting as a heat sink for

design basis events but are not appropriate or applicable during the

refueling or cold shutdown conditions. No changes have been made to

the actual suppression pool temperature or level limits and

therefore, the assumptions made in the accident and transient

analyses remain valid. These limits are consistent with the STS. The

revisions to the surveillance requirements are also intended to

improve clarity and consistency with STS. The deletion of the

requirement to monitor suppression pool water temperature every 5

minutes during relief valve operation is appropriate in that plant

operating and emergency operating procedures already specify what

actions are to be taken when suppression pool average water

temperature increases above 95 deg.F including initiation of

suppression pool cooling. Monitoring pool temperature every 5

minutes during these events is not necessary and is redundant to

other actions. Therefore, these changes will not significantly

increase the probability of occurrence of the consequences of an

accident previously evaluated.

Containment Atmosphere Dilution

The revisions to the applicability of TS section 3.7.H,

``Containment Atmosphere Dilution'', requiring the containment

atmosphere dilution system to be operable only when the reactor is

in power operation and the primary containment is required to be

inerted will not significantly increase the probability or

consequences of an accident previously evaluated because the CAD

system can only function when the containment is inerted. The

function of the CAD system is to inject nitrogen into the

containment after a LOCA and ensure the containment remains inerted.

Drywell inspections performed after plant startup and prior to plant

shutdown require that the primary containment be de-inerted for

personnel access. Therefore, CAD system operability is not required

during these inspections. No changes to the actual function or

purpose of the CAD system are proposed.

Oxygen Concentration

The changes to TS section 3/4.7.1, ``Oxygen Concentration'' are

administrative in that they only clarify the requirement that both

the suppression chamber and the drywell must have oxygen

concentrations less than 4 [percent] by volume. The revisions to the

surveillance requirements are consistent with STS. Decreasing the

frequency of verification of oxygen concentration from twice per

week to once per week is in accordance with STS and reflects the

fact that during power operation, the containment is inerted and

slightly pressurized such that air (oxygen) cannot leak into the

containment. Therefore, these changes will not significantly

increase the probability or consequences of an accident previously

evaluated.

Secondary Containment

The deletion of the requirement to operate the SGTS immediately

after a secondary containment violation is identified will not

affect the reliability of the secondary containment in that

containment integrity is normally fully restored immediately after a

violation is identified. The testing of the SGTS involves insertion

of a Group III containment isolation signal and is only appropriate

if the restoration of secondary containment involves a temporary or

new secondary containment boundary. These modifications to a

secondary containment boundary, however, would require that the SGTS

be operated as part of post modification testing. Deleting the

requirement for the SGTS to be operated after minor secondary

containment violations will reduce the possibility of human error

(such as misalignment of valves after the system is returned to its

normal configuration) due to reduced testing. Operation of the SGTS

after a secondary containment violation is not required by STS.

Revision of the definition of calm wind conditions will not

affect the reliability or availability of the secondary containment

or SGTS. An engineering evaluation on the effects of wind speed and

direction on the ability of the SGTS to maintain 1/4'' vacuum in

secondary containment has been performed. The results indicate that

while wind effects can be seen on individual instruments, there is

minimal effect on the average instrument readings with wind speeds

up to 15 mph. A discussion of this evaluation has been added to the

Bases of TS section 3.7. Therefore, these changes will not

significantly increase the probability or the consequences of an

accident previously evaluated.

Secondary Containment Automatic Isolation Dampers

The addition of operability requirements, actions and

surveillance requirements for secondary containment isolation

dampers better ensures the integrity and isolation capability of the

secondary containment. The new specifications are consistent with

the requirements of the STS. The actual function or operation of the

secondary containment isolation valves/dampers will not be affected.

The appropriate valves/dampers will be incorporated in plant

procedures that are subject to the change control provisions of TS.

Therefore, these changes will not increase the probability of

occurrence or consequences of an accident previously evaluated in

the TS.

Standby Gas Treatment System

The change to the output requirements of the inlet heaters for

each train of the SGTS from 11 kw to 22 kw better ensures that these

heaters (and the SGTS) can perform their design function. The 22 kw

output requirement ensures that the inlet air humidity does not

exceed the 70 [percent] humidity specified in the UFSAR. This change

does not affect the actual operation of the heaters or the SGTS.

The requirement to demonstrate the HEPA filter uniform air

distribution after HEPA filter replacement or after structural

maintenance on the filter system housing (rather than annually) will

not decrease the reliability of the SGTS. The air flow test will be

performed after work or modifications which have the ability to

disrupt the system geometry or result in potential flow blockage.

Revising the shutdown LCO requirement in the various

specifications from requiring the plant to be in Cold Shutdown in 24

hours to requiring Hot Shutdown in 12 hours and Cold Shutdown (or

other condition not requiring equipment operability) in the

following 24 hours is consistent with STS and the shutdown

requirements in TS section 3.5. This new requirement will allow the

reactor to be shutdown in a more controlled manner and will not

result in a significant increase in the probability or consequences

of an accident previously evaluated.

The revisions to the Bases are administrative in that they only

reflect the changes to the individual specifications described

previously in this section or correct minor discrepancies. All

changes are consistent with the applicable specifications.

(2) The proposed amendment will not increase the possibility of

a new or different kind of accident from any accident previously

evaluated for the following reasons.

As described in the above response to question 1, none

of the proposed changes alters the design of the plant or equipment

or the plant's transient response. The changes to the definitions

and limiting conditions for operation applicable to TS section 3.7

are consistent with STS and better ensure that equipment assumed to

be operable in our accident analysis will be operable upon demand.

The addition of limiting conditions for operation for drywell

average temperature and secondary containment isolation valves will

better ensure that the assumptions in our accident analysis remain

valid.

The changes to the surveillance requirements are consistent with

the STS. Those systems required to mitigate accidents evaluated in

the UFSAR will still be operable and available.

The reduction in conditional surveillance testing of certain

systems and equipment will reduce the probability of equipment

failure as a result of excessive testing or due to human error.

(3) The proposed amendment will not involve a significant

reduction in a margin of safety for the following reasons.

The revisions to the limiting conditions for operation in

Chapter 3.7 of the TS will not invalidate the original licensing

basis assumptions and will not invalidate any assumptions or input

parameters for any DAEC event analysis. These changes provide more

specific guidance only and are in accordance with the STS. Extending

the time period within which the DAEC must achieve Cold Shutdown

conditions will permit increased operator attention and minimal

distractions for operators during shutdown, thus minimizing the

risks of unexpected operational transients.

Additional surveillance testing for certain instrumentation and

systems will provide additional assurance that these systems will be

available when needed.

Elimination of unnecessary or conditional surveillance testing

will not reduce the minimum necessary equipment operability

requirements or equipment reliability. Elimination of the redundant

testing will reduce equipment failure due to excessive testing or

human error.

In summary, the proposed administrative changes do not change

the probability or consequences of an accident previously evaluated,

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

and do not involve a reduction in the margin of safety.

Therefore, the proposed license amendment is judged to involve

no significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

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

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

Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC

20036

NRC Project Director: John N. Hannon

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

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

Connecticut

Date of amendment request: May 27, 1974

Description of amendment request: The amendment would temporarily

allow the Operations Manager to not have a senior reactor operator

(SRO) license for Millstone 3, providing other conditions are met.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:...The proposed change does not

involve an SHC [significant hazards consideration] because the change

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change affects only an administrative control,

which was based on the existing industry guidance in ANSI N18.1-

1971, that recommended the Operations Manager hold a SRO license.

The current guidance in Section 4.2.2 of ANSI/ANS-3.1-1987

recommends, as one option, that the Operations Manager have held a

license for a similar unit and the Operations Middle Manager hold a

SRO license. The Operations Middle Manager position does not exist

at NNECO [Northeast Nuclear Energy Company]. Therefore, the proposed

change requests an exception to ANSI N18.1-1971 to allow use of

ANSI/ANS-3.101987 in a limited circumstance. Specifically, the

proposed revision to Technical Specification 6.3.1 would temporarily

allow the Operations Manager to have held a SRO at a PWR

[pressurized water reactor] other than Millstone Unit No. 3. The

proposed revision would be in effect for the period ending three

years after the Staff's approval for this request.

The proposed exception to ANSI N18.1-1971 will allow at least

one of the Operations Assistants (instead of a Operations Middle

Manager) to hold, and continue to hold, a SRO license, if the

Operations Manager does not hold a license. The proposed change

includes the requirement if the Operations Manager does not hold a

SRO license at Millstone Unit No. 3, he shall have held a license

for a similar unit in accordance with Section 4.2.2 of ANSI/ANS-3.1-

1987. For those areas of knowledge that require a SRO license, at

least one of the Operations Assistants hold a SRO license and

provides technical guidance normally required by the Operations

Manager.

The proposed change does not alter the design of any system,

structure, or component. It does not change the way any plant

systems are operated. It does not reduce the knowledge,

qualifications, or skills of any operator on watch, and does not

affect the way the Operations Department is managed by the

Operations Manager in maintaining the effective performance of his

personnel and to ensure the plant is operated safely and in

accordance with the requirements of the Operating License.

The proposed change does not detract from the Operations

Manager's ability to perform his primary responsibilities. In this

case, by having previously held a SRO license for a similar unit, he

will have gained the necessary training, skills, and experience to

fully understand the operation of plant equipment and the watch

requirements for operators.

The proposed change does not weaken the supervisory chain that

presently exists in the Operations Department. All Control Room

operators will continue to be supervised by the licensed Shift

Supervisor.

In summary, the proposed change does not affect the ability of

the Operations Manager to provide the plant oversight required of

his position. In addition, it does not have any [e]ffect on the

probability or consequences of any previously evaluated accident.

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

from any previously evaluated.

The proposed change to Technical Specification 6.3.1 does not

affect the design or function of any plant system, structure, or

component. It does not affect, in any way, the performance of NRC

licensed operators, nor does it change the way any plant equipment

is operated. Operation of the plant in conformance with technical

specifications and other license requirements will continue to be

supervised by personnel who hold an NRC SRO license. The proposed

change to Technical Specification 6.3.1 ensures that the Operations

Manager will be a knowledgeable and qualified individual. The

proposed change does not introduce any new failure modes.

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

The proposed change involves only an administrative control

which is not related to the margin of safety as defined in the

technical specifications. The proposed change does not reduce the

level of knowledge or experience required of an individual who fills

the Operations Manager position, nor does it affect the conservative

manner in which the plant is operated. All Control Room operators

will continue to be supervised by personnel who hold a SRO license.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

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

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

NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: May 31, 1994

Description of amendment request: This amendment will change the

frequency for monitoring the Spray Pond ground water level from once

per month to once every six months in Technnical Specification

Requirement 4.7.1.3.c for each Susquehanna unit.

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:

I. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed Technical Specification change to extend the

monthly surveillance interval for Spray Pond ground water level

measurement to biannual does not affect the the probability or

consequences of an accident previously evaluated. The safety

analysis performed for this change concludes that the ground water

level is stable, predictable, and significantly below the acceptance

criteria established in the Technical Specifications. Thus, less

frequent monitoring of the ground water level does not increase the

probability that the Spray Pond will become inoperable due to rising

ground water levels or the probability of any accident scenarios

associated with Spray Pond inoperability. The Technical

Specification change will not impact the function or the method of

operation of plant systems, structures, or components. Thus, the

consequences of a malfunction of equipment important to safety

previously evaluated in the FSAR are not increased by the change.

II. This proposal does not create the possibility of a new or

different kind of accident or from any accident previously

evaluated.

The proposed Technical Specification change to extend the

monthly surveillance interval for Spray Pond ground water level

measurement to biannual does not create the possibility of a new or

different kind of accident or from any accident previously

evaluated. The proposed change dies not affect systems, structures,

or components (SSCs) or the operation of the SSCs; and therefore

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

accident.

III. This change does not involve a significant reduction in a

margin of safety.

The proposed Technical Specification change to extend the

monthly surveillance frequency to biannual does not reduce the

margin of safety. The ground water level in the vicinity of the

Spray Pond has been proven to be stable and predictable through

twelve years of monthly data collection. This data has shown the

highest ground water levels (still considerably lower than the

Technical Specification limit) to occur in the months of April and

October. Therefore, surveillance of the ground water level at the

observation sites during April and October will adequately monitor

this aspect of the Spray Pond operability.

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: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: Charles L. Miller

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

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: May 25, 1994

Description of amendment request: Salem is in the process of

upgrading the Radiation Monitoring System. This upgrade will involve a

replacement on many of the existing radiation monitors. The proposed

change modifies Tables associated with Technical Specifications 3/

4.3.3.1 Radiation Monitoring Instrumentation, 3/4.3.3.8 Radioactive

Liquid Effluent Monitoring Instrumentation, and 3/4.3.3.9 Radioactive

Gaseous Effluent Monitoring Instrumentation. The proposed change

relocates the Salem specific radiation monitor numbers from the table

to a cross reference in the Bases. No required radiation monitoring

functions are being changed.

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 proposal does not involve] a significant increase in

the probability or consequences of an accident previously analyzed.

The proposed change to relocate the Salem specific radiation

monitor numbers to the Bases is administrative. There are no

modifications or changes in operating conditions associated with the

proposed changes. Providing a more accurate description of a

referenced radiation monitor in the note is editorial. The proposed

changes do not affect the probability of occurrence or the

consequences of accidents identified in the UFSAR. No accident

precursors are being generated by the proposed changes. Therefore,

the proposed changes do not involve a significant increase in the

probability or consequences of a previously analyzed accident.

2. [This proposal does not create] the possibility of a new or

different kind of accident.

The proposed changes to relocate the Salem specific radiation

monitor numbers to the Bases are administrative. The change to the

note on Table 3.3-6 is editorial to provide a more accurate

description of the required radiation monitor. There are no

modifications or changes in operating conditions associated with the

proposed changes. Therefore, the proposed changes will not increase

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

accident previously identified.

3. [These changes do not involve] a significant reduction in a

margin of safety.

The Technical Specification operability requirements for the

radiation monitors are not being changed. Relocating the Salem

specific radiation monitor numbers to the Bases will not change any

requirements for the radiation monitors. The change to the note on

Table 3.3-6 is editorial to provide a more accurate description of

the required radiation monitor. Therefore, the changes to the

surveillance frequencies do not involve a significant reduction in

any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: Charles L. Miller

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: May 23, 1994, as supplemented June 15,

1994

Description of amendment request: The proposed amendment would

revise the Ginna Station Technical Specification (TS) 3.1.4 regarding

allowable primary coolant levels of specific activity. The limit for I-

131 dose equivalent of iodine activity in the reactor coolant would be

increased from 0.2 to 1.0 micro Ci/gm. The limit for total specific

activity of the reactor coolant would be increased from 84 to 100/E

micro Ci/gm, where E is the average beta and gamma energies per

disintegration in Mev. Both increased allowable levels are consistent

with NUREG-1431 ``Standard Technical Specifications, Westinghouse

Plants, September 1992,'' and NUREG-0800 ``Standard Review Plan for the

Review of Safety Analysis Reports for Nuclear Power Plants'' (Section

15.6.3).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed changes do not affect any accident initiators and

therefore the probability of any accident is not increased.

Consequences of the changes are analyzed and shown acceptable in the

[Westinghouse LOFTTR2 Analysis of Potential Radiological

Consequences Following a Steam Generator Tube Rupture at the R.E.

Ginna Nuclear Power Plant] analysis, WCAP-11668, Section III.

2. Operation of Ginna Station in accordance with the proposed

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

of accident from any accident previously evaluated.

The proposed changes involve no physical modifications to the

plant; therefore, no new accident can be postulated.

3. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant reduction in a margin of

safety, as no margin of safety is reduced by the proposed changes,

as shown in WCAP-11668.

Based upon the above information, it has been determined that

the proposed changes to the Ginna Station Technical Specifications

do not involve a significant increase in the probability or

consequences of an accident previously evaluated, does not create

the possibility of a new or different kind of accident previously

evaluated, and does not involve a significant reduction in a margin

of safety. Therefore, it is concluded that the proposed changes meet

the requirements of 10 CFR 50.92(c) and do not involve a significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005

NRC Project Director: Walter R. Butler

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

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

Date of amendment request: June 6, 1994

Description of amendment request: The proposed amendment would

merge Toledo Edison Company into Cleveland Electric Illuminating

Company. As described in the application, the company formed from the

merger is intended to be renamed. Therefore, the licensee uses the

nomenclature ``NEWCO'' as a temporary name of the combined operating

company, and will provide the permanent name by supplemental letter.

The amendment would (1) replace the Toledo Edison Company and Cleveland

Electric Illuminating Company with ``NEWCO'' as a licensee, (2)

designate ``NEWCO'' as the owner of the Davis-Besse Nuclear Power

Station, Unit 1, and (3) make other associated changes to the license

as indicated in the amendment application. Centerior Service Company

would be unaffected by the amendment and would remain a licensee.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below, indicating that the proposed

changes would:

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

accident previously evaluated because no accident initiators or

assumptions are affected. The proposed changes are administrative

and have no direct affect on any plant systems. All Limiting

Conditions for Operation, Limiting Safety System Settings and Safety

Limits specified in the Technical Specifications will remain

unchanged.

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

accident previously evaluated because no accident conditions or

assumptions are affected. The proposed changes do not alter the

source term, containment isolation, or allowable radiological

consequences. The proposed changes are administrative and have no

direct effect on any plant systems.

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

any accident previously evaluated because no new accident initiators

are created. The proposed changes are administrative and have no

direct effect on any plant systems. The changes do not affect the

reactor coolant system pressure boundary and do not affect any

system functional requirements, plant maintenance, or operability

requirements.

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

from any accident previously evaluated because no different accident

initiators are created. The proposed changes are administrative and

have no direct effect on any plant systems. The changes do not

affect the reactor coolant system pressure boundary and do not

affect any system functional requirements, plant maintenance, or

operability requirements.

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

because the proposed changes do not involve new or significant

changes to the initial conditions contributing to accident severity

or consequences. The proposed changes are administrative and have no

direct affect on any plant systems.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo Library,

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

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

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

NRC Project Director: John N. Hannon

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: May 20, 1994

Description of amendment request: The proposed amendment would

remove Core Spray (CS) High Sparger Pressure Instrumentation from the

Vermont Yankee Technical Specifications for Emergency Core Cooling

System (ECCS) Actuation Instrumentation. In addition, an unrelated

administrative change is also proposed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change to remove the Core Spray High Sparger

Pressure Instrumentation from the Technical Specifications for ECCS

Actuation Instrumentation is consistent with NRC requirements

concerning this instrumentation. This instrumentation is considered

NNS [nonnuclear safety] and performs a local monitoring and alarm

function only. In addition, the NRC has recently approved the

removal of Core Spray Sparger Break Detection Instrumentation from

the Technical Specifications of another BWR [boiling water reactor]

with a similar situation.

The CS Sparger Piping is inspected every refueling outage to

verify its integrity. No cracks in the CS Sparger piping have been

identified since the first inspection in 1980. CS Sparger Piping

integrity is still assured. The instrumentation systems to be

removed from the ECCS Actuation Instrumentation Technical

Specifications do not perform any automatic control or trip

function. In addition, this instrumentation does not provide

information that is required to permit the control room operator to

take manual actions that are required for safety systems to

accomplish their safety functions for design basis accident events.

The proposed change does not result in any system hardware

modification, function change or new plant configuration. The

requested change to ECCS Actuation Instrumentation does not impact

any FSAR [Final Safety Analysis Report] safety analysis involving

the ECCS or Protection Systems. These monitoring functions are not

contributors to the initiation of accidents.

The administrative changes to correct typographical errors on

Tables 3.2.1 and 4.2.1 will have no affect on plant hardware, plant

design, safety limit setting or plant system operation and

therefore, do not modify or add any initiating parameters that would

significantly increase the probability or consequences of any

previously analyzed accident.

Therefore, it is concluded that there is not a significant

increase in the probability or consequence of an accident previously

evaluated.

2. The function of the Core Spray High Sparger Pressure

Instrumentation to be removed from the Technical Specifications is

for local indication and alarm only. These functions are not

necessary for operators to accomplish any safety functions.

The proposed change does not involve any change in hardware,

function, Technical Specification trip setpoints, plant operation,

redundancy, protective function or design basis of the plant. There

is no impact on any existing safety analysis or safety design

limits. Core Spray High Sparger Pressure Instrumentation functions

do not initiate nuclear system parameter variations which are

considered potential initiating causes of threats to the fuel and

the nuclear system process barrier.

As discussed above, the proposed administrative change only

corrects typographical errors concerning equipment identification

numbers. This change doe not affect any equipment and it does not

involve any potential initiating events that would create any new or

different kind of accident.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change to remove the Core Spray High Sparger

Pressure Instrumentation from the Technical Specifications for ECCS

Actuation Instrumentation does not affect any existing safety

margins. This equipment is NNS and performs a local indication and

alarming function only. The original intent of this detection system

was because the first BWR plants had only the CS System for long-

term core cooling. Later, plants like VY [Vermont Yankee] were

provided with Low Pressure Coolant Injection (LPCI) Systems in

addition to CS.

Existing Technical Specification requirements for automatic trip

functions are unaffected. Failure of the Core Spray High Sparger

Pressure Instrumentation does not preclude the ability of the CS

System to perform its safety function to mitigate the consequences

of accidents or of any other safety system to accomplish its safety

functions. Proper ECCS functioning post-accident is not relied upon

by NNS alarming functions but by such systems as safety related

reactor level indication.

The CS Sparger Piping is inspected every refueling outage to

verify its integrity. No cracks in the CS Sparger piping have been

identified since the first inspection in 1980. The removal from the

Technical Specifications has no affect on the bases of Protective

Instrumentation which is to operate to initiate required system

protective actions. The Core Spray High Sparger Pressure

Instrumentation does not perform any safety function.

As discussed above, the proposed administrative change which

corrects typographical errors does not affect any equipment involved

in potential initiating events or safety limits. [***].

Based upon the above, it is concluded that the proposed changes

do not involve a significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont 05301

Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,

One International Place, Boston, Massachusetts 02110-2624

NRC Project Director: Walter R. Butler

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

Project No. 2, Benton County, Washington

Date of amendment request: March 18, 1992, modifications submitted

June 25 and July 28, 1992, and December 6, 1993

Description of amendment request: On December 6, 1993, the licensee

submitted a significant modification to its original request of March

18, 1992 (April 19, 1992 (57 FR 12349)). The modified request would

amend the Technical Specifications (TS) to provide two surveillance

tests to determine the operability of the catalyst beds in the

containment atmospheric control (CAC) system. One test compares

hydrogen content in the influent to the hydrogen content in effluent

process streams to assure the catalyst is operating. The second test

measures the temperature profile in the catalyst bed to ensure that

sufficient catalyst remains available for the recombination process

during postulated accident conditions. Since the original proposal has

been significantly changed, the staff is issuing a new notice and

proposed no significant hazards consideration determination.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change does not involve any changes in the design

or operation of the hydrogen recombiners, which are accident

mitigation systems. The proposed change revises surveillance

requirements to ensure that existing equipment will perform as

designed in response to postulated events. The proposed change does

not, therefore, involve an increase in the probability or

consequences of an accident previously evaluated.

2. Does the amendment create the possibility of a new or

different kind of accident from any accident previously evaluated?

The proposed change does not involve any changes in the design

or operation of existing equipment, and does not, therefore, create

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

accident previously evaluated.

3. Does the amendment involve a significant reduction in a

margin of safety?

The proposed change increases surveillance requirements for

existing equipment to ensure that the equipment will perform as

designed. With equipment performing as designed in response to

postulated accidents, the proposed change does not affect any

existing margins of safety.

Based on the licensee's analysis and the staff's analysis, it

appears that the three standards of 10 CFR 50.92(c) are satisfied.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

Attorney for licensee: Nicholas S. Reynolds, Esq., Winston &

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

NRC Project Director: Theodore R. Quay

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: March 31, 1994

Description of amendment request: The proposed amendment would

revise the Kewaunee Nuclear Power Plant Technical Specifications (TS)

by incorporating operability and surveillance requirements for the

recently installed Auxiliary Feedwater Pump Low Discharge Pressure Trip

instrumentation. Proposed surveillance requirements would be added to

Table TS 4.1-1, ``Minimum Frequencies for Checks, Calibrations and Test

of Instrument Channels.'' TS 3.4, ``Steam and Power Conversions

System,'' would be revised to explicitly link operability of the

associated Auxiliary Feedwater Pump Low Discharge Pressure Trip channel

to operability of the associated auxiliary feedwater pump. In addition,

minor format inconsistencies in TS 3.4.b.1.A and 3.4.b.1.B would be

corrected.

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) Table TS 4.1-1

The proposed changes were reviewed in accordance with the

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

proposed changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change defines the necessary surveillance

requirements for the recently installed Auxiliary Feedwater Pump Low

Discharge Pressure Trip channels. The intent of adding surveillance

requirements to the TS's is to ensure the availability and

reliability of the components. The proposed change is an additional

restriction not presently included in the TS's. Therefore, it will

not increase the probability or consequences of an accident

previously evaluated in the USAR.

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

from any accident previously evaluated.

The proposed change adds surveillance requirements to the TS for

the Auxiliary Feedwater Low Discharge Pressure Trip channels. It

does not alter the plant configuration or overall plant performance.

Therefore, it does not create the possibility of a new or different

kind of accident.

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

This proposed revision is an additional requirement in the TS's

to ensure the availability and reliability of the Auxiliary

Feedwater Pump Low Discharge Pressure Trip channels. It does not

alter the input or assumptions of the safety analysis, and is an

enhancement from an overall safety standpoint. Therefore, it will

not involve a reduction in the margin of safety.

(b) TS 3.4

The proposed changes were reviewed in accordance with the

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

proposed changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change defines the necessary operability

requirements for the recently installed Auxiliary Feedwater Pump Low

Discharge Pressure Trip channels. Installation of this protection

was recommended and approved by the NRC prior to their installation.

The proposed change requires that the reactor not be heated

350 deg.F unless both motor driven Auxiliary Feedwater

Pumps and their associated low discharge pressure trip channels are

operable. Also, the reactor shall not be heated

350 deg.F unless the turbine driven auxiliary feedwater pump and its

associated low discharge pressure trip channel are operable, or if

not demonstrated operable prior to 350 deg.F, they

shall be declared inoperable when 350 deg.F is exceeded.

Furthermore, when the reactor is 350 deg.F, an auxiliary

feedwater pump low discharge pressure trip channel may be inoperable

for a period not to exceed 4 hours. If this time is exceeded, the

associated auxiliary feedwater pump shall be declared inoperable and

the appropriate limiting condition for operation of TS 3.4.b.2

entered. The intent of adding these operability requirements to the

TS's is to ensure the availability of the components. The proposed

change is an additional restriction not presently included in the

TS's. Therefore, it will not increase the probability or

consequences of an accident previously evaluated in the USAR.

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

from any accident previously evaluated.

The proposed change adds operability requirements to the TS for

the Auxiliary Feedwater Pump Low Discharge Pressure Trip channels.

It does not alter the plant configuration or overall plant

performance. Therefore, it does not create the possibility of a new

or different kind of accident.

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

This proposed revision is an additional requirement in the TS's

to ensure the operability of the Auxiliary Feedwater Pump Low

Discharge Pressure Trip channels. It does not alter the input or

assumptions of the safety analysis, and is an enhancement from an

overall safety standpoint. Therefore it will not involve a reduction

in the margin of safety

(c) Administrative changes to TS 3.4.b.1.A and TS 3.4.b.1.B

The proposed changes were reviewed in accordance with the

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

proposed changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, or

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

from an accident previously evaluated, or

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

The proposed changes are administrative in nature and do not

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

significant hazards exist.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Wisconsin

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

54301.

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

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

NRC Project Director: John N. Hannon

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: May 24, 1994

Description of amendment request: The amendment request proposes to

revise the Technical Specifications (TS) to implement the NRC's Final

Policy Statement on Technical Specification Improvements for Nuclear

Power Reactors (58 FR 39132). These improvements involve focusing the

Technical Specifications on those requirements that are of controlling

importance to operational safety by screening each TS in Section 3/4.1

through 3/4.11 using the criteria provided in the policy statement. The

purpose of the proposed amendment request is to relocate the

specifications that do not meet any of the four policy statement

criteria. The relocated specifications will be moved to Updated Final

Safety Analysis (USAR) Chapter 16. Based on the screening, all or part

of 38 technical specifications were identified as not meeting any of

the criteria and, therefore, as candidates for relocation. The licensee

has categorized the TS changes as (1) specifications relocated intact

to USAR Chapter 16, (2) specifications relocated with portions retained

in TS, (3) specifications relocated with programmatic requirements

referenced in Section 6 of TS, (4) modifications to retained

specifications to accomodate relocation of other specifications, and

(5) new specification requirements incorporated into the TS. The last

category is used to effect the retention of portions of relocated

specifications and accomodate the policy statement recommendation to

incorporate industry experience in the determination of TS content.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed Technical Specification changes involve relocating

requirements that are not conditions or limitations on reactor

operation necessary to obviate the possibility of an abnormal

situation or event giving rise to an immediate threat to the public

health and safety. The proposed changes were identified through the

application of criteria designed to cull those requirements that are

not important to operational safety from the Technical

Specifications. In this process, selected provisions of the

Technical Specifications identified for relocation were retained if

necessary to support a Technical Specification that was to be

retained. Thus, only specification requirements that have little or

no operational safety significance are proposed for relocation. In

addition, those requirements that would be relocated will be

included in the Updated Final Safety Analysis Report (USAR) and,

therefore, will be controlled and implemented as NRC commitments. In

this manner, those requirements that have no operational safety

significance but involve maintaining the plant in its as-designed

state (for example, through surveillance programs) would be

controlled.

In addition, the criteria for identifying requirements to be

retained in the Technical Specifications specifically call out, for

retention, those structures, systems, or components that are

required to mitigate accidents previously evaluated.

Based on the above, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

different kind of accident from any previously evaluated.

The proposed changes involve relocating Technical Specification

requirements to another licensee-controlled document. No changes or

physical alterations of the plant are involved. Also, no changes to

the operation of the plant or equipment are involved. Therefore, the

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

kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in

the margin of safety.

The proposed changes involve relocating Technical Specification

requirements to the USAR. The requirements to be relocated were

identified by applying the criteria endorsed in the Commission's

Policy Statement. Thus, those specifications that would be relocated

do not impose constraints on design and operation of the plant that

are derived form the plant safety analysis report or from

probabilistic safety assessment (PSA) information and do not belong

in the Technical Specifications in accordance with 10 CFR 50.36 and

the purpose of the Technical Specifications stated in the Policy

Statement. Therefore, relocation of these requirements does not

involve a significant reduction in the margin of safety.

In addition, revisions to the USAR will be evaluated in

accordance with the 10 CFR 50.59 process which considers the

reduction in safety margin. Therefore, any future revisions to the

provisions in the USAR will consider reductions in the margin of

safety using the criteria for identifying an unreviewed safety

question.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

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

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

66621

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

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

NRC Project Director: Theodore T. Quay

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: June 7, 1994

Description of amendment request: The proposed amendment revises

Technical Specification Table 2.2-1, Reactor Trip System

Instrumentation Setpoints, to change the over-temperature-delta-

temperature (OTDT) axial flux difference (AFD) limits to reflect

results of the Cycle 8 core maneuvering analysis.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The probability of occurrence and the consequences of an

accident evaluated previously in the Updated Safety Analysis Report

(USAR) are not increased due to the proposed technical specification

change. Operation at 3565 MWt does not affect any of the mechanisms

postulated in the USAR to cause LOCA or non-LOCA design basis

events. Analyses, evaluations and minimum DNBR [departure from

nucleate boiling ratio] calculations confirm that the USAR

conclusions remain valid for the proposed changes. On these bases it

is concluded that the probability and consequences of the accidents

previously evaluated in the USAR are not increased.

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

or different kind of accident from any previously evaluated.

There is no new type of accident or malfunction being created.

The proposed change provides revised operating limits necessary to

support Cycle 8, and does not change the method and manner of plant

operation. The safety design bases in the USAR have not been

altered. Thus, this change does not create the possibility of a new

or different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The proposed changes do not change the plant configuration in a

way that introduces a new potential hazard to the plant and do not

involve a significant reduction in the margin of safety. The

analyses and evaluations discussed in the safety evaluation

demonstrate that all applicable safety analysis acceptance criteria

continue to be met for the proposed operating conditions. Items not

specifically cited in this safety evaluation have been reviewed and

have been found to be bounded by the evaluations performed for

Reference 1 [Wolf Creek Generating Station Technical

Specifications]. Therefore, it is concluded that the margin of

safety, as described in the bases to any technical specification, is

not reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

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

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

66621

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

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

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.

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

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

Date of amendment request: May 5, 1994, as supplemented June 16,

1994.

Description of amendment request: The proposed amendments would

change the Technical Specifications to increase Main Steam and

Pressurizer Code Safety Valve Setpoint Tolerances.

Date of publication of individual notice in Federal Register: June

21, 1994 (59 FR 32029).

Expiration date of individual notice: July 21, 1994

Local Public Document Room location:

Atkins Library, University of North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

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.

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 29, 1992, as

supplemented on October 22, 1993, and November 11, 1993.

Brief description of amendments: The amendments revise the Site

Boundary Map and the Low Population Zone Map.

Date of issuance: June 22, 1994

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

30 days.

Amendment Nos.: 190 and 167

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 28, 1992 (57 FR

48813) The Commission prepared an Environmental Assessment and Finding

of No Significant Impact which was published in the Federal Register on

May 13, 1994 (59 FR 25129). The Commission's related evaluation of

these amendments is contained in a Safety Evaluation dated June 22,

1994.No significant hazards consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

Station,Plymouth County, Massachusetts

Date of application for amendment: February 11, 1993, as

supplemented December 2, 1993, January 5, February 22, March 1, April

15, and May 16, 1994.

Brief description of amendment: This amendment increases the

allowed fuel assembly storage cells from 2320 to 3859, changes the

maximum loads allowed to travel over the spent fuel assemblies from

1050 to 2000 lbs., and changes the limiting characteristics of

assemblies to be stored in the spent fuel from a maximum KINIFITY

less than or equal to 1.35 to a Maximum KINIFITY less than or

equal to 1.32 and a maximum lattice average uranium enrichment of less

than or equal to 4.6% by weight.

Date of issuance: June 22, 1994

Effective date: June 22, 1994

Amendment No.: 155

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

Technical Specifications.

Date of initial notice in Federal Register: April 30, 1993 (58 FR

26171) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 22, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

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

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

Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1

and 2, Rock Island County, IllinoisDate of application for

amendments: March 26, 1993

Brief description of amendments: The amendments revise Technical

Specification 3/4.6 for Dresden and Quad Cities Stations to allow

Single Loop Operation (SLO) with the recirculation loop suction and

discharge valves open. The amendments also delete outdated and

unnecessary portions of Technical Specification 3.6.H for Dresden Units

2 and 3 and provide more consistency to the BWR Standard Technical

Specifications (NUREG-0123, Revision 4).

Date of issuance: June 16, 1994

Effective date: June 16, 1994

Amendment Nos.: 127, 121, 147, and 143

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

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 13, 1994 (59 FR

17594) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 16, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: For Dresden, Morris Public

Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,

Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.

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

Nuclear Station, Units 1 and 2, York County, South CarolinaDate of

application for amendments: March 24, 1994, as supplemented April

11 and May 31, 1994

Brief description of amendments: The amendments revise the

Technical Specification (TS) to increase boron concentration for the

spent fuel storage pool during Modes 1-3 operation and for the

refueling canal during Mode 6 operation; include two reload related

topical reports in TS 6.9.1.9; and correct errors in nomenclature and

remove obsolete footnotes.

Date of issuance: June 13, 1994

Effective date: June 13, 1994

Amendment Nos.: 120 and 114

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 28, 1994 (59 FR

22006) The April 11 and May 31, 1994, letters provided clarifying and

additional information that did not change the scope of the March 24,

1994, application and the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated June 13, 1994.No

significant hazards consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

ConnecticutDate of application for amendment: December 17, 1993, as

supplemented April 12, 1994.

Brief description of amendment: The amendment changes the action

statements for the limiting conditions for operation associated with

the electrical power sources (Technical Specification 3.8.1.1).

Date of issuance: June 14, 1994

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

30 days.

Amendment No.: 177

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

Technical Specifications.

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

4943). The April 12, 1994, letter provided clarifying information that

did not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated June 14, 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.

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

A. FitzPatrick Nuclear Power Plant, Oswego County, New YorkDate of

application for amendment: December 28, 1993

Brief description of amendment: The amendment revises Technical

Specification (TS) Section 6.9(A)1.a. to permit startup reports for

cycles subsequent to the initial fuel cycle to address only those

startup tests that are actually performed. The amendment also revises

TS Section 6.9(A) to clarify requirements for the submission of routine

reports. These changes are consistent with the guidance provided in

NUREG-1433, ``Standard Technical Specifications - General Electric

Plants, BWR/4.''

Date of issuance: June 16, 1994

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

30 days.

Amendment No.: 212

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

Technical Specifications.

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

4945) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 16, 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. 50-333, James

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: December 29, 1993

Brief description of amendment: The amendment revises Appendix B of

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

Specifically, the amendment revises Appendix B Surveillance Requirement

3.1.a. and Table 3.10-2 to provide surveillance requirements for data

recorders associated with the gaseous effluent monitoring system. The

amendment also makes an editorial change to Appendix B Limiting

Condition for Operation 3.1.a. to improve consistency and clarity.

Date of issuance: June 16, 1994Effective date: As of the date of

issuance to be implemented within 30 days.

Amendment No.: 213

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

Technical Specifications.

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

4946) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 16, 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.

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

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama.

Date of amendments request: May 13, 1991, as supplemented October

13, 1992.

Brief description of amendments: The amendments modify the TS for

the overpressure protection systems. The allowable outage time (AOT)

for one inoperable residual heat removal (RHR) relief valve with one or

more of the reactor coolant system cold leg temperatures less than or

equal to 310 degrees Fahrenheit is being decreased from 7 days to 24

hours for water-solid conditions. The required AOT for low temperature

conditions, other than water-solid, will remain at 7 days with one RHR

relief valve inoperable, provided the pressurizer level is less than or

equal to 30 percent and a dedicated operator is assigned to monitor and

control the reactor coolant system pressure.

Date of issuance: June 16, 1994

Effective date: June 16, 1994

Amendment Nos.: 108 and 100

Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise

the Technical Specifications.

Date of initial notice in Federal Register: July 22, 1992 (57 FR

32577) and February 17, 1993 (58 FR 8787)The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

June 16, 1994.No significant hazards consideration comments received:

No

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

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

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

Ferry Nuclear Plant, Units 1 and 3, Limestone County, Alabama

Date of application for amendment: April 1, 1992 (TS 302)

Brief description of amendments: The amendments add requirements to

the Browns Ferry Units 1 and 3 Technical Specifications to provide

administrative controls for a post-accident sampling system, which were

requested by Generic Letter 83-36, ``NUREG-0737 Technical

Specifications.''

Date of issuance: June 21, 1994

Effective date: June 21, 1994

Amendment Nos.:207 and 180

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

Date of initial notice in Federal Register: May 27, 1992 (57 FR

22269) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 21, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

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: December 23, 1992, as

supplemented on March 18, 1994.

Brief description of amendment: This amendment revises TS 3/4 3.3.5

for transfer switches used to meet 10 CFR Part 50, Appendix R (Fire

Protection) requirements, and specifies a new special report

requirement for TS 6.9.2.

Date of issuance: June 14, 1994

Effective date: June 14, 1994

Amendment No. 187

Facility Operating License No. NPF-3. The amendment revised the

Technical Specifications.

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

10016) The supplemental information submitted on March 18, 1994, did

not change the initial proposed finding of no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated June 14, 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.

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

Project No. 2, Benton County, Washington

Date of application for amendment: December 20, 1993, as amended

March 25 and April 25, 1994

Brief description of amendment: The amendment modifies the

technical specifications (TS) to address new containment purge and vent

valves to be installed in the 1994 refueling outage. The amendment

changes the containment purge and vent valve TS as follows: (1) removes

the requirement ensuring that valve position remains at less than or

equal to 70 degrees, (2) changes the containment leak testing

requirements for the metal-to-metal seated valves from 6 months to 2

years since they have improved seat designs, and (3) makes

administrative changes to delete an out-of-date note, to relocate an

action statement requirement from the TS surveillance section to the TS

action statement section, and to change a related TS reference to this

surveillance section. Valve opening position does not need to be

limited to less than or equal to 70 degrees. The resiliently-seated

valves have a permanently installed mechanical stop to limit the open

position to ensure adequate closure times. The metal-to-metal seated

valves are designed to close from the 90-degree open position.

Date of issuance: June 15, 1994

Effective date: 15 days from the date of issuance

Amendment No.: 124

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

Technical Specifications.

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

14901) The additional information contained in the March 25 and April

25, 1994, letters was clarifying in nature, is 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 amendment is contained in a Safety Evaluation

dated June 15, 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

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

Project No. 2, Benton County, Washington

Date of application for amendment: February 8, 1994, as

supplemented March 25, 1994

Brief description of amendment: The amendment revises the WNP-2

Technical Specifications. Specifically, the amendment increases 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 deletes

the Note (j) reference from RCIC-V-8 and RCIC-V-63. Note (j) indicates

that the stroke time specified in the table reflects the requirement

for containment isolation only.

Date of issuance: June 17, 1994

Effective date: June 17, 1994

Amendment No.: 125

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

Technical Specifications.

Date of initial notice in Federal Register: May 12, 1994 (59 FR

24754) The additional information contained in the March 25, 1994,

letter was clarifying in nature, was 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 amendment is contained in a Safety Evaluation dated

June 17, 1994.No significant hazards consideration comments received:

No.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352]

Dated at Rockville, Maryland, this 28th day of June 1994.

FOR THE NUCLEAR REGULATORY COMMISSION

Jack W. Roe,

Director, Division of Reactor Projects - III/IVOffice of Nuclear

Reactor Regulation

[Doc 94-16174 Filed 7-5-94 8:45 am]

BILLING CODE 7590-01-F