Biweekly Notice

[[Page 65672]]

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from November 27, 1995, through December 8, 1995.

The last biweekly notice was published on December 6, 1995 (60 FR

62485).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

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

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

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

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

intervene is discussed below.

By January 19, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any

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limitations in the order granting leave to intervene, and have the

opportunity to participate fully in the conduct of the hearing,

including the opportunity to present evidence and cross-examine

witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555, Attention: Docketing and Services

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

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

Nuclear Regulatory Commission, Washington, DC 20555, and to the

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

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

Date of amendments request: November 7, 1995

Description of amendments request: The proposed amendment would

adopt the improved Standard Technical Specifications (NUREG-1432)

format and content of Section 5.0, ``Design Features,'' as modified by

approved changes to the improved Standard Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The Proposed amendment does not change the Design Features, only

relocates the information to other documents. This is consistent

with the NRC Policy Statement and NUREG-1432. Therefore, relocating

existing information, eliminating information which duplicates

information found in other licensee documents, and making

administrative improvements provide Technical Specifications which

are easier to use. Because information is relocated to established

programs where changes to those programs are controlled by

regulatory requirements, there is no reduction in commitment and

adequate control is still maintained. Likewise, the elimination of

information which duplicates information in other licensee

documents, enhances the useability of the Technical Specifications

without reducing commitments. The administrative improvements being

proposed neither add nor delete requirements, but merely clarify and

improve the understanding and readability of the Technical

Specifications. Since the requirements remain the same, these

changes only affect the method of presentation and are considered

administrative, and as such, would not affect possible initiating

events for accidents previously evaluated or any system functional

requirement.

Therefore, the proposed changes would not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The relocation of existing requirements, the elimination of

requirements which duplicate existing information, and making

administrative improvements are all changes that are administrative

in nature. The proposed changes will not affect any plant system or

structure, nor will they affect any system functional or operability

requirements. Consequently, no new failure modes are introduced as a

result of the proposed changes. The proposed changes are consistent

with the improved Standard Technical Specifications, for the most

part, as plant specific information is included in this section.

Therefore, the proposed change would not create the possibility of a

new or different type of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes are administrative in nature in that no

change[s] to the design features of the facility are being made. The

Design Features Section is being reformatted to be consistent, for

the most part, with NUREG-1432, ``Standard Technical Specifications,

Combustion Engineering Plants,'' Revision 1. The proposed changes do

not affect the UFSAR design bases, accident assumptions, or

Technical Specification Bases. In addition, the proposed changes do

not affect release limits, monitoring equipment, or practices.

Consequently, the proposed changes would not involve a significant

reduction in a margin of safety.

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

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

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

amendments request involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

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

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

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: William H. Bateman

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County, Connecticut

Date of amendment request: October 20, 1995

Description of amendment request: The proposed amendment would

revise the Electrical Power Systems Surveillance Intervals from 18

months to once per refueling (i.e., nominal 24 months).

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

[[Page 65674]]

below. The no significant hazards consideration analysis has been

divided into three parts: AC Sources Operating, DC Sources Operating,

and On-Site Power distribution:

In accordance with 10CFR50.92, CYAPCO has reviewed the proposed

changes and concluded that they do not involve an SHC. The basis for

this conclusion is that the three criteria of 10CFR50.92(c) are not

compromised. The proposed changes do not involve an SHC because the

changes would not:

AC Sources Operating

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change will increase the interval between a

surveillance that is performed during plant shutdown from once per

18 months to a maximum of once per 30 months (i.e., 24 months

nominal + 25% as allowed by Specification 4.0.2). The proposed

change to Surveillance Requirement 4.8.1.1.2.f does not alter the

intent or the method by which the surveillance is conducted. In

addition, the acceptance criterion for the surveillance is

unchanged. As such, the proposed change will not degrade the ability

of the EDG [emergency diesel generator] to perform its intended

function.

A review of the past surveillances, and preventive maintenance

of the diesel generators indicates that the appropriate acceptance

criterion was met in each case. Additional assurance of the diesel

generator's operability is provided by Surveillance Requirement

4.8.1.1.2.a.4 and the performance of other on-line testing as

described above. As such, the proposed changes do not adversely

affect the probability of an accident previously analyzed.

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

than any accident previously evaluated.

The proposed change regarding the testing frequency of the

diesel generators [i.e., from once per 18 months to a maximum of

once per 30 months (i.e., 24 months + 25 percent as allowed by

Specification 4.0.2)] does not affect the operation or response of

any plant equipment, including the diesel generators, or introduce

any new failure mechanism. The proposed change does not affect the

test acceptance criteria of the EDGs. The plant equipment will

respond per design and analyses, and there will not be a malfunction

of a new or different type introduced by the testing frequency

revision to the EDG surveillance requirements. As such, the changes

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

from any previously analyzed.

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

The Bases Section of Technical Specification Section 3/4.8,

``Electrical Power Systems,'' states that the operability of the AC

and DC power systems and associated distribution systems ensure that

sufficient power will be available to supply the safety-related

equipment required for safe shutdown and mitigation and control of

accident conditions. Bases Section 3/4.8 also states that the

surveillance requirements for determining the operability of the

EDGs are in accordance with the recommendations of Regulatory Guide

1.108, Revision 1. The revision of surveillance requirements will

continue to verify that the EDGs are operable. Operable EDGs ensure

that the assumptions in the Bases of the Technical Specifications

are not affected and ensure that the margin of safety is not

reduced. Therefore, the assumptions in the Bases of the Technical

Specifications are not affected and the change does not result in a

significant reduction in the margin of safety.

DC Sources Operating

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

CYAPCO is proposing to modify the frequency of Surveillance

Requirements 4.8.2.1.c, d, and f of the Haddam Neck Plant Technical

Specifications from at least once per 18 months to at least once

each refueling interval. These surveillance requirements verify the

operability of components of the Class 1E DC power system. CYAPCO is

also proposing to delete the term ``during shutdown'' contained in

Surveillance Requirements 4.8.2.1.d, 4.8.2.1.e, and 4.8.2.1.f.

Additional assurance of the operability of the Class 1E DC power

system is provided by Surveillance Requirements 4.8.2.1.a, b, and e.

The proposed changes do not alter the intent or method by which

the surveillances are conducted, do not involve any physical changes

to the plant, do not alter the way any structure, system, or

component functions, and do not modify the manner in which the plant

is operated. As such, the proposed changes in the frequency of

Surveillance Requirements 4.8.2.1.c, d, and f will not degrade the

ability of the Class 1E DC power system to perform its intended

safety function. Also, the Class 1E DC power system is designed to

perform its intended safety function even in the event of a single

failure.

Equipment performance over the last four operating cycles was

evaluated to determine the impact of extending the frequency of

Surveillance Requirements 4.8.2.1.c, d and f. cThis evaluation

included a review of surveillance results, preventive maintenance

associated with normal surveillance activities, and corrective

maintenance records. It concluded that the Class 1E DC power system

is highly reliable, and that there is no indication that the

proposed extension could cause deterioration in the condition or

performance of any of the subject Class 1E DC power system

components.

The deletion of the phrase ``during shutdown'' in Surveillance

Requirement 4.8.2.1.d, e, and f is acceptable. The terms ``Cold

Shutdown'' and ``Hot Shutdown'' are defined in the Haddam Neck Plant

Technical Specifications as operating modes or conditions. The

proposed deletion of the term ``during shutdown'' is intended to

prevent possible misinterpretations and is consistent with the

recommendations of GL 91-04.

Based on the above, the proposed changes to Surveillance

Requirements 4.8.2.1.c, d, e, and f of the Haddam Neck Plant

Technical Specifications do not involve a significant increase in

the probability or consequences of an accident previously analyzed.

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

than any accident previously evaluated.

CYAPCO is proposing to modify the frequency of Surveillance

Requirements 4.8.2.1.c, d, and f of the Haddam Neck Plant Technical

Specifications from at least once per 18 months to at least once

each refueling interval. CYAPCO is also proposing to delete the term

``during shutdown'' contained in Surveillance Requirements

4.8.2.1.d, 4.8.2.1.e, and 4.8.2.1.f. These surveillance requirements

verify the operability of components of the Class 1E DC power

system.

The proposed changes do not alter the intent or method by which

the surveillances are conducted, do not involve any physical changes

to the plant, do not alter the way any structure, system, or

component functions, and do not modify the manner in which the plant

is operated. As such, the proposed changes to Surveillance

Requirements 4.8.2.1.c, d, e, and f will not introduce a new failure

mode.

Based on the above, the proposed changes to Surveillance

Requirements 4.8.2.1.c, d, e, and f of the Haddam Neck Plant

Technical Specifications will not create the possibility of a new or

different kind of accident from any previously evaluated.

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

CYAPCO is proposing to modify the frequency of Surveillance

Requirements 4.8.2.1.c, d, and f of the Haddam Neck Plant Technical

Specifications from at least once per 18 months to at least once

each refueling interval. CYAPCO is also proposing to delete the term

``during shutdown'' contained in Surveillance Requirements

4.8.2.1.d, 4.8.2.1.e, and 4.8.2.1.f. These surveillance requirements

verify the operability of components of the Class 1E DC power

system.

Equipment performance over the last four operating cycles was

evaluated to determine the impact of extending the frequency of

Surveillance Requirements 4.8.2.1.c, d and f. This evaluation

included a review of surveillance results, preventive maintenance

associated with normal surveillance activities, and corrective

maintenance records. It concluded that the Class 1E DC power system

is highly reliable, and that there is no indication that the

proposed extension could cause deterioration in the condition or

performance of any of the subject Class 1E DC power system

components.

Additional assurance of the operability of the Class 1E DC power

system is provided by Surveillance Requirements 4.8.2.1.a, b, and e.

Since decreasing the surveillance frequency does not involve a

significant increase in the consequences of a design basis accident

previously analyzed, the proposed changes to Surveillance

Requirements 4.8.2.1.c, d, e, and f of the Haddam Neck Plant

Technical Specifications do not involve a significant reduction in

the margin of safety.

[[Page 65675]]

On-Site Power Distribution

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to Surveillance Requirement 4.8.3.1.2 will

increase the surveillance interval from once each refueling outage

(once per 18 months) to a maximum of once per 30 months (i.e., 24

months nominal + 25% as allowed by Specification 4.0.2.). The

proposed change to Surveillance Requirement 4.8.3.1.2 does not alter

the intent or the method by which the surveillance is conducted. In

addition, the acceptance criterion for the surveillance is

unchanged. As such, the proposed changes will not degrade the

ability of the MCC-5 ABT scheme to perform its intended function.

The successful past surveillance results, and the simpler re-

design of the MCC-5 ABT provide assurance of system operability up

to a maximum of 30 months. As such, the proposed changes do not

adversely affect the probability or consequences of an accident

previously analyzed.

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

than any accident previously evaluated.

The proposed change does not alter the intent or method by which

the surveillance is conducted, does not involve any physical changes

to the plant, does not alter the way any structure, system, or

component functions, and does not modify the manner in which the

plant is operated. As such, the proposed change to Surveillance

Requirement 4.8.3.1.2 will not introduce a new failure mode.

Based on the above, the proposed change to Surveillance

Requirement 4.8.3.1.3 of the Haddam Neck Plant Technical

Specifications will not create the possibility of a new or different

kind of accident from any previously evaluated.

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

The proposed change to Surveillance Requirement 4.8.3.1.2

extends the frequency for verifying the operability of the MCC-5 ABT

scheme from at least once per 18 months to at least once per

refueling interval (i.e., 24 months nominal + 25% as allowed by

Specification 4.0.2).

The proposed change does not alter the intent or method by which

the surveillance is conducted, does not involve any physical changes

to the plant, does not alter the way any structure, system, or

component functions, and does not modify the manner in which the

plant is operated. As such, the proposed change in the frequency of

Surveillance Requirement 4.8.3.1.2 will not degrade the ability of

the MCC-5 ABT to perform its safety function and does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270.

NRC Project Director: Phillip F. McKee

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County, Connecticut

Date of amendment request: October 27, 1995

Description of amendment request: The proposed amendment will

revise Technical Specification (TS) Section 3.6.3, ``Containment

Isolation Valves.'' These changes will clarify the action statement for

when a penetration has only one containment isolation valve (CIV) and

that valve is inoperable.

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 (SHC), which is presented below:

...The proposed change does not involve an SHC because the

change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The containment isolation system is an engineered safety feature

that functions to allow normal or emergency passage of fluids

through the containment boundary, while preserving the ability of

the boundary to prevent or limit the escape of fission products that

may result from postulated accidents.

All fluid system pipelines that penetrate the containment are

provided with one or more valves that can be closed remotely, either

electrically or pneumatically, or are locked manual valves. Most of

the piping penetrations connect to equipment inside the reactor

containment. Thus, they are not open to the reactor containment

atmosphere and will not pass radioactive contamination to the CIV

unless the pipe is ruptured inside containment during an accident.

Lines that penetrate the reactor containment and are not in

service during operation are isolated with one or more locked closed

CIVs. Lines that are in service and that pass fluids during

operation are provided with one or more motor-operated valves,

positive closure trip valves, or check-valves.

The lack of guidance contained in Technical Specification

Section 3.6.3 for a penetration that has only one CIV in it, does

not increase the probability or consequences of an accident

previously evaluated. This design, and the consequences that could

result from this configuration have been evaluated previously and

found acceptable. The proposed modification simply provides guidance

to the operators should a penetration with only one CIV becomes

inoperable. This proposed technical specification will, as do other

technical specification action statements, provide a reasonable time

to correct the situation before a required shutdown must commence.

In addition, this proposed Action Statement was developed to be

consistent with Technical Specification Section 3.0.3.

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

from any accident previously evaluated.

The proposed modification provides guidance to the operators

should a penetration which has only one CIV be inoperable. This

design has been previously evaluated and found to be acceptable from

both a deterministic and probabilistic standpoint. The proposed

modification will provide the operators specific guidance to restore

the penetration to an operable state or to isolate it. With this

guidance, they can avert the risk associated with a plant shutdown,

which would be mandated without this guidance. Should a CIV be

inoperable and not capable of being restored, the proposed technical

specification provides additional options. However, a probabilistic

risk assessment review has determined that these additional options

are not risk significant. Finally, the containment isolation system

cannot be an accident initiator, rather it is designed to respond to

accidents. The inability of the CIVs to operate cannot create a new

or different kind of accident from any previously evaluated.

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

The proposed modification provides the requirement to the

operators should a penetration which has only one CIV be inoperable.

The effects of this design have been previously evaluated and found

to be acceptable from both a deterministic and probabilistic

standpoint.

The current Haddam Neck Plant containment isolation system has

been previously reviewed by the NRC. CYAPCO is not making any

changes to the containment isolation system. CYAPCO is however,

providing guidance in the technical specifications should a

penetration which has only one CIV be inoperable. This guidance will

allow CYAPCO to correct the event associated with the penetration

with an NRC approved alternative, in a set time. This provision is

safe especially when compared to the alternative which is a plant

shutdown under Technical Specification Section 3.0.3.

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: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270.

[[Page 65676]]

NRC Project Director: Phillip F. McKee

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

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

Date of amendment request: August 17, 1995

Description of amendment request: The Commission issued Amendment

Nos. 128 and 122 to the Facility Operating Licenses for Catawba Units 1

and 2 on February 17, 1995, which revised Technical Specification (TS)

Table 2.2-1 and TS Surveillance Requirement (SR) 4.2.5 to allow a

change in the method for measuring reactor coolant system (RCS)

flowrate from the calorimetric heat balance method to a method based on

a one-time calibration of the RCS cold leg elbow differential pressure

taps. In its application submitted on January 10, 1994, for the above

listed amendments, Duke Power (the licensee) neglected to modify SR

4.2.5.2 to delete that portion of the SR that specifies that the

measurement instrumentation shall be calibrated within 7 days prior to

the performance of the flowrate measurement. The licensee states that

the requirement to calibrate the measurement instrumentation within 7

days prior to the performance of the flowrate measurement is

impractical based on utilization of the cold leg elbow pressure tap

method of RCS flowrate measurement. Accordingly, the licensee proposes

to modify SR 4.2.5.2 to reflect the deletion of the subject

requirement.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. This change is considered administrative in nature and

should have been requested in Duke Power Company's January 10, 1994

application, as amended. The instrumentation which was subject to

the requirement is no longer utilized in the fulfillment of the TS

required RCS flowrate determination. The proposed changes will not

result in any impact upon accident probabilities, since the RCS

flowrate measurement instrumentation is not accident initiating

equipment. Likewise, they will not result in any impact upon

accident consequences, since no change to any method or frequency of

calibration of the RCS flowrate transmitters will result. The plant

response to accidents will not be affected.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. No change is being made to any plant design feature, or

to the manner in which the plant will be operated. Therefore, no new

accident causal mechanisms can be generated. As noted above, the

proposed changes are considered administrative in nature, and should

have been requested in the January 10, 1994 application, as amended.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. No impact upon any fission product

barriers will occur as a result of the approval of the proposed

changes. No change to plant design, operating, maintenance, or test

characteristics will result from the proposed amendments. No impact

upon any plant safety margins will result.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: November 15, 1995

Description of amendment request: The proposed amendments modify

Technical Specification (TS) 3/4.7.1 and the associated Bases to

increase the setpoint tolerance of the main steam safety valves (MSSVs)

from plus or minus one percent to plus or minus three percent, to

incorporate a requirement to reset as-left MSSV lift settings to within

plus or minus one percent following surveillance testing, and to delete

two obsolete footnotes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. As demonstrated previously, all applicable licensing

basis safety analyses were evaluated with a MSSV setpoint drift of

plus or minus 3%. The results of the evaluations were within all

appropriate accident analysis acceptance criteria. No significant

impact on DNBR results, peak primary or secondary pressures, peak

fuel cladding temperature, dose, or any other accident analysis

acceptance criterion was involved. No impact on the probability of

any accident occurring exists as a result of the increased MSSV

setpoint tolerance.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. No change is being made to any plant design feature, or

to the manner in which the plant will be operated. Therefore, no new

accident causal mechanisms can be generated. The MSSV setpoint

tolerance only affects the time at which the valve opens following

or during a transient, and is not a contributor to the probability

of an accident.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. As stated above, all relevant

accident analyses were examined to determine the effect of the wider

MSSV setpoint tolerances. All analysis results are within applicable

acceptance criteria. Finally, the NRC has previously approved TS

changes for other plants seeking to use the [plus or minus] 3

[percent] setpoint tolerance, including McGuire Nuclear Station

(reference Amendment Nos. 146 and 128 for Units 1 and 2,

respectively).

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: November 15, 1995

Description of amendment request: The proposed amendments modify

Technical Specification (TS) Limiting Condition for Operation 3.7.5 to

raise the minimum nuclear service water system's (RN) water level in

the standby nuclear service water pond (SNSWP) from 570 to 571 feet

mean sea level.

[[Page 65677]]

This change will increase the volume of water that will be available

for use of the SNSWP as the ultimate heat sink for postulated accidents

under all meteorological conditions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The requested amendments will not involve a significant increase

in the probabililty or consequences of an accident previously

evaluated. The proposed amendments will have no impact upon any

accident probabilities, since the RN system is not a accident

initiating system. It is an accident mitigating system. Accident

consequences will not be affected, since the proposed amendments

will require a greater surface area for heat transfer from the SNSWP

water to the environment. It has been determined that with the

required TS minimum water level of 571 feet and with the required TS

temperature limit of 91.5F [degrees Farenheit], the SNSWP will be

capable of fulfilling all design basis requirements pertaining to

accident mitigation.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. As stated previously, the RN system is not an accident

initiator. No change is being made to the plant which would cause

the RN system to become an accident initiator. All relevant

procedures will be changed as required, commensurate with the NRC

issuance of the requested amendments. No accident causal mechanisms

will be affected. The effect of the increased SNSWP level on the

SNSWP dam was evaluated and found to be negligible.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. As noted above, the SNSWP was

evaluatd with the new TS level requirement and was determined to be

operable and capable of meeting all design basis requirements. No

impact on any fission product barriers is created by the proposed

changes. The proposed changes will ensure that the RN system remains

capable of fulfilling its required accident mitigating functions.

SNSWP temperature will continue to be monitored at an elevation of

568 feet, which is considered to be the highest elevation at which

the average SNSWP surface temperature is accurately represented and

minimally influenced by daily temperature swings due to variations

in solar heat input, air temperature, and rainfall temperature.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: August 20, 1992, as supplemented

December 5, 1995.

Description of amendment request: The proposed amendments, would

revise the Technical Specifications (TS) related to the 60-month 120-

volt battery surveillance requirement. The proposed change is to delete

the words ``during shutdown'' from SR 4.8.2.1.2.e (performance

discharge test). The licensee contends that the ``during shutdown''

provision in the TS is an impractical requirement because both units

would have to be shutdown to perform the performance discharge test

(PDT).

In the licensee's supplement dated December 5, 1995, proposed

changes were made to TS 3/4 8.2 Bases to support the frequency of the

PDT on the other batteries in the system after a battery that had its

PDT performed is returned to service.

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 seeks to change the surveillance

requirements to allow the performance with the units on line. The

surveillance can be safely completed as proposed without affecting

unit operation. The equipment would not be removed from service for

a time that would exceed the current allowed outage time. The

probability or consequences of any accident previously evaluated

will not be increased because the removal of a battery from service

can be performed while on line, and the loads of each battery can be

assumed by another same-train battery which is the case for the

battery being inoperable for any other reason. During the allowed

outage time, even a single failure of any component (including

Emergency Diesel Generator) will still leave a full capacity train

available to provide instrumentation and control power for both

units. Train redundancy is maintained at all times. Compensatory

action is taken to prohibit discharge testing of the other remaining

batteries within 10 days following a battery performance discharge

test to ensure that the tested battery is fully recharged.

Probabilistic Risk Analysis shows that the increase in Core Damage

Frequency due to this operation is negligible.

2. The proposed amendment will not change any actual

surveillance requirements, the change would simply allow the

requirements to be met at different unit conditions. The performance

of the surveillance with the units on line does not require any new

component configurations that would reduce the ability of any

equipment to mitigate an accident. The station would not be in any

degraded status beyond that which has previously been evaluated.

Therefore the proposed change will not create the possibility of a

new accident.

3. The change would allow a battery to be removed from service

for testing. However, the testing must be completed within the

current allowed outage time. As the allowed outage time defines the

required margin of safety for equipment operability, removing

equipment from service for testing and returning it to service

within the allowed time does not affect a margin of safety.

Compensatory action is taken to prohibit discharge testing of the

other remaining batteries within 10 days following a battery

performance discharge test to ensure that the tested battery is

fully recharged.

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: Atkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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: November 6, 1995

Description of amendment request: The proposed amendment would

revise the alarm setpoints for the noble gas and in-containment high

range area radiation monitors listed in Table 3.3-6 of Beaver Valley

Power Station, Unit 1 Technical Specification (TS) 3.3.3.1. The

proposed revisions would make these alarm setpoints consistent with the

criteria in the Emergency Action Levels (EALs) which were revised and

[[Page 65678]]

approved by the NRC in August 1994. The revised EALs use the noble gas

radiation monitors as indications of effluent releases and are based on

dose to the public. The revised EALs use the in-containment high range

area radiation monitors as indication of fission product barrier

challenges or failures rather than as indications of effluent release.

The proposed amendment would also revise Action Statement 36 of

Table 3.3-6 of TS 3.3.3.1 for both BVPS-1 and BVPS-2 to reflect a

previously approved change in reporting frequency for effluent

releases. BVPS-1 License Amendment No. 188 and BVPS-2 License Amendment

No. 70 (both issued on June 12, 1995) approved a change in the

reporting frequency for effluent releases from semi-annual to annual.

The proposed change would make Action Statement 36 consistent with this

previously approved change.

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 monitor alarm setpoint changes and editorial

changes are administrative in nature. Should the radiation alarm

fail to annunciate or give a false alarm, there would be no affect

on any other plant equipment or systems. The noble gas monitors are

not safety related and do not interface with any safety related

system. The containment area monitors are safety related; however,

they do not initiate any safety function, nor do they interface with

any other safety related system.

The monitors' alarm as a visual (lighted icon) and audible alarm

in the control room. The operator is then responsible for taking any

corrective actions necessary, based on the alarm and Emergency

Action Level (EAL) guidelines. The monitors do not provide for any

automatic actions of other equipment or systems when an alarm

condition occurs.

The operating and design parameters of the radiation monitors

will not change. The proposed change affects only the radiation

level at which an alarm condition is created and does not affect any

accident assumptions or radiological consequences of an accident.

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 proposed radiation monitor alarm revisions cannot initiate a

new type of accident. A failure of the monitor itself cannot serve

as the initiating event of an accident and has no effect on the

operation of a safety system. Operator action is not made solely on

a radiation monitor alarm; other plant condition indicators are also

evaluated.

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 referenced radiation monitoring channels have no capability

to mitigate the consequences of an accident. Also, they do not

interface with any safety related system. The containment area

monitors are safety related channels which provide indication to the

operator of the integrity of the fission product barriers in

containment. This indication, combined with other indications of

plant conditions may direct an operator to take action to mitigate

the consequences of an accident. The alarm setpoint itself does not

perform any specific safety related function and the trip value is

not referenced in the Updated Final Safety Analysis Report (UFSAR),

nor does any site design basis document take credit for this

setpoint. Safety limits and limiting safety system settings are not

affected by this proposed change. Also, the site will continue to

meet the requirements of 10 CFR Part 100 which limits offsite dose

following a postulated fission product release.

Therefore, use of the proposed technical specification would not

involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

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

NRC Project Director: John F. Stolz

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: November 22, 1995

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications (TS) Index to

delete reference to the BASES. The proposed revisions to Turkey Point

Units 3 and 4 TS are administrative in nature. Changes to the TS BASES

will be controlled by a plant procedure under administrative controls

and reviews. Proposed changes to the TS BASES will be evaluated in

accordance with 10 CFR 50.59.

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments are administrative in nature and do not

affect assumptions contained in plant safety analyses, the physical

design and operation of the plant, nor do they affect Technical

Specifications that preserve safety analysis assumptions. The

Technical Specification BASES, per 10 CFR 50.36(a), are not a part

of the Technical Specifications. Changes to the TS BASES will be

controlled by a plant procedure under administrative controls and

reviews. Proposed changes to the TS BASES will be evaluated in

accordance with 10 CFR 50.59. Therefore, the proposed change does

not affect the probability or consequences of accidents previously

analyzed.

(2) The proposed license amendments do not create the

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

previously evaluated.

The proposed amendments are administrative in nature. The

proposed amendments will not create the possibility of a new or

different kind of accident from any accident previously evaluated

since the proposed amendments will not change the physical plant or

the modes of plant operation defined in the facility operating

license. No new failure mode is introduced due to the administrative

change, since the proposed change does not involve the addition or

modification of equipment nor does it alter the design or operation

of affected plant systems, structures, or components.

(3) The proposed license amendments do not involve a significant

reduction in a margin of safety.

The operating limits and functional capabilities of the affected

systems, structures, and components are unchanged by the proposed

amendments. The BASES information, per 10 CFR 50.36(a), is not a

part of the Technical Specifications. Changes to the TS BASES will

be controlled by a plant procedure under administrative controls and

reviews. Proposed changes to the TS BASES will be evaluated in

accordance with 10 CFR 50.59. Therefore, the proposed change does

not reduce 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 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: Florida International

[[Page 65679]]

University, University Park, Miami, Florida 33199

Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036

NRC Project Director: David B. Matthews

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: November 3, 1995

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) to delay for one cycle the

volumetric and surface examinations of the Reactor Coolant Pump (RCP)

motor flywheels required by Regulatory Guide (RG) 1.14, Regulatory

position C.4.b, incorporated by reference in Technical Specification

5.6.2.8.c, to coincide with Crystal River Unit 3 (CR-3) Refueling

Outage 11, scheduled for Spring 1998.

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 will not significantly increase the

probability or consequences of an accident previously evaluated.

The safety function of the RCP flywheels is to provide a

coastdown period during which the RCPs would continue to provide

reactor coolant flow to the reactor after loss of power to the RCPs.

The maximum loading on the RCP motor flywheel results from overspeed

following a large LOCA [loss-of-coolant accident]. The estimated

maximum obtainable speed in the event of a Reactor Coolant System

piping break was established conservatively. The proposed one time

change does not affect that analysis. Reduced coastdown times due to

a single failed flywheel would not place the plant in an unanalyzed

condition since a locked rotor (instantaneous coastdown) is analyzed

in the FSAR [Final Safety Analysis Report]. The proposed change does

not increase the amount of radioactive material available for

release or modify any systems used for mitigation of such releases

during accident conditions. Therefore, the proposed change does not

involve a significant increase in the consequences of any accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change will not change the design, configuration,

or method of operation of the plant. Therefore, the proposed change

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

accident from any previously evaluated.

3. The proposed change will not involve a significant reduction

to any margin of safety.

FPC [Florida Power Corporation] has performed two full

volumetric examinations in excess of those recommended in RG 1.14,

Revision 1 during the Second ISI [inservice inspection] Interval.

The margins of safety defined in RG 1.14, Revision 1 used in the

analysis are not significantly changed.

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: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

Attorney for licensee: A. H. Stephens, General Counsel, Florida

Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida

33733

NRC Project Director: David B. Matthews

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,

Appling County, Georgia

Date of amendment request: November 10, 1995

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) for containment systems to

reflect the adoption of the requirements of 10 CFR Part 50, Appendix J,

Option B, and the implementation of a performance-based containment

leak-rate testing program at the Edwin I. Hatch Nuclear Plant, Units 1

and 2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

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

the probability of consequences of an accident previously evaluated.

The proposed changes do not involve any physical or operational

changes to structures, systems or components. The proposed changes

provide a mechanism within the TS for implementing a performance-

based leakage rate test program which was promulgated by the

revision to 10 CFR 50 to incorporate Option B to Appendix J. The TS

Limiting Conditions for Operation (LCO) remain unaffected by these

changes. Thus, the safety design basis for the accident mitigation

functions of the primary containment, the airlocks, and the primary

containment isolation valves is maintained. Therefore, these changes

will not increase the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously analyzed.

Revising Surveillance Requirement acceptance criteria and

frequencies does not physically modify the plant and does not modify

the operation of any existing equipment.

3. The proposed changes do not involve a significant reduction

in the margin of safety, nor do they affect a safety limit, an LCO,

or the manner in which plant equipment is operated. The NRC letter

dated November 2, 1995, recognizes that changes similar to the

proposed changes are required to implement Option B of 10 CFR 50,

Appendix J. In NUREG-1493, ``Performance-Based Containment Leak-Test

Program,'' which forms the basis for the Appendix J revision, the

NRC concludes that adoption of performance-based test intervals for

Appendix J testing will not significantly reduce the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location:Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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

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

NRC Project Director: Herbert N. Berkow

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island

Nuclear Station, Unit No. 2 (TMI-2), Dauphin County, Pennsylvania

Date of amendment request: January 16, 1995

Description of amendment request: The proposed amendment would

revise TMI-2 Operating License No. DPR-73 by modifying Section 6.5.1.7

of the administrative controls portion of the technical specifications.

The revision would change Section 6.5.1.7 to delete the requirement for

personnel in the internal GPU Nuclear (GPUN) Review and Approval matrix

to render an unreviewed safety question (USQ) determination regarding

(1) proposed changes to unit technical specifications and (2)

investigations of violations of technical specifications. Both of these

activities involve docketed correspondence with the NRC in which the

USQ determination is made and justified. This obviates the need for a

requirement for the licensee to perform and document an internal USQ

[[Page 65680]]

determination. This change would make the TMI-2 Technical

Specifications consistent with the Standard Technical Specifications

for B&W Plants (NUREG 1430).

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:

10 CFR 50.92 provides the criteria which the Commission uses to

perform a no significant hazards consideration. 10 CFR 50.92 states

that an amendment to a facility license involves no significant

hazards if 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 proposed change to the technical specifications is

administrative and does not involve any physical changes to the

facility. No changes are made to operating limits or parameters, nor

to any surveillance activities. Based on this, GPU Nuclear has

concluded that the proposed change does not:

1. Involve a significant increase in the probability of

occurrence of the consequences of an accident previously evaluated.

The proposed amendment is purely administrative and affects only the

review of activities that involve considerable review by the NRC.

This change will not degrade the performance of review for either of

the two activities that are affected. This proposed technical

specification change does not involve changes to hardware

configuration, operation, or testing. Therefore, this change does

not increase the probability of occurrence or the consequences of an

accident previously evaluated.

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

since the change is administrative and no new failure modes are

created.

3. Involve a change in the margin of safety. This change is

administrative in nature; compatible with standard technical

specifications; and does not affect any safety settings, equipment,

or operational parameters.

Based on the above analysis it is concluded that the proposed

changes involve no significant safety hazards considerations as

defined by 10 CFR 50.92.

The NRC staff has reviewed the analysis of the licensee 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: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105

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

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

NRC Project Director: Seymour H. Weiss

Illinois Power Company and Soyland Power Cooperative, Inc., Docket

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

Illinois

Date of amendment request: October 27, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.1.3, ``Control Rod OPERABILITY,''

to include the 25% surveillance overrun allowed by Limiting Condition

for Operation (LCO) 3.0.2 into the allowances of the surveillance Notes

for control rod ``notch'' testing per Surveillance Requirement (SR)

3.1.3.2 and SR 3.1.3.3. The proposal also includes a clarification to

the description of TS Table 3.3.3.1-1, ``Post Accident Monitoring

Instrumentation,'' Function 7, to indicate that the Function's

requirements apply to the position indication for only automatic

primary containment isolation valves, rather than all primary

containment isolation valves. Finally, the proposal includes changes to

correct a number of editorial and typographical errors inadvertently

contained in TS 3.3.4.1, ``End of Cycle Recirculation Pump Trip (EOC-

RPT) Instrumentation,'' TS 3.3.6.1, ``Primary Containment and Drywell

Isolation Instrumentation,'' TS 3.3.8.2, ``Reactor Protection System

(RPS) Electric Power Monitoring,'' and TS 3.6.5.2, ``Drywell Air

Lock.''

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 associated with Limiting Condition for

Operation (LCO) 3.1.3 are being made to make the surveillance

requirement (SR) Notes agree with their original intent. The Notes

were originally intended to allow the testing of control rods to be

tracked as a group, i.e., partially withdrawn and fully withdrawn.

In the event that a control rod(s) has changed from one test group

to another, the Notes were intended to allow performance of the next

surveillance on that control rod(s) to be delayed to coincide with

the next regularly scheduled performance of the test of the new

group. However, these Notes failed to include the 25% surveillance

extension allowances of SR 3.0.2. This proposed change merely adds

the 25% extension to the time allowed by the Notes to make them

agree with the Frequency plus the extension allowance of SR 3.0.2.

The addition of the word ``fully'' to the Note for SR 3.1.3.2 is to

provide for clarification only. These changes are consistent with

changes approved for the Grand Gulf Nuclear Station (GGNS) and River

Bend Station and are being proposed for the Clinton Power Station

(CPS) for consistency. The proposed changes do not involve a change

to the control rods or control rod drive system design or operation.

Further, the proposed change does not affect the way in which the

associated control rod test is performed, only the ``triggers'' for

performance of the test are affected. These triggers are being

revised to make them consistent with their original intent. As a

result, the proposed change cannot increase the probability or the

consequences of any accident previously evaluated.

The proposed change to the description of LCO 3.3.3.1 Function 7

to include ``automatic'' is provided for clarification only. As

described in the Bases for this Function, the requirements for

operability are currently only associated with automatic primary

containment isolation valves (PCIVs). As a result, this change does

not involve a change to the scope of this LCO. In addition, these

changes are consistent with changes approved for GGNS and are being

proposed for CPS for consistency. Since this request does not affect

the design or operation of this equipment, nor does it alter the

scope of this Technical Specification (TS) requirement, this

proposed change cannot increase the probability or the consequences

or any accident previously evaluated.

The remaining proposed changes are purely editorial and do not

affect the design or operation of any equipment or alter the

technical requirements of any TS. As a result, these proposed

changes cannot increase the probability or the consequences of any

accident previously evaluated.

(2) The proposed changes do not affect the design or operation

of any equipment. In addition, the proposed changes do not affect

the manner in which any test is performed or involve a change to any

plant operating mode or configuration. As a result, Illinois Power

has concluded that the proposed changes cannot create the

possibility of an accident not previously evaluated.

(3) The proposed changes to the SRs for LCO 3.1.3 are being made

to make the SR Notes agree with their original intent and thus

permit control rods to be tested as originally intended. The

proposed changes do not involve a change to the control rods or

control rod drive system design or operation. Further, the proposed

change does not affect the way in which this test is performed or

the routine Frequency of performing the test, only the ``triggers''

are affected. Since these triggers are being revised to make them

consistent with their original intent, Illinois Power has determined

that this change does not result in a reduction in the margin of

safety.

The proposed change to the description of LCO 3.3.3.1 Function 7

to include ``automatic'' is provided for clarification only. As

described in the Bases for this Function, the requirements for

operability are

[[Page 65681]]

currently only associated with automatic PCIVs. As a result, this

change does not involve a change to the current scope of this LCO.

Since this request does not affect the design or operation of this

equipment, nor does it alter the scope of this TS requirement, this

proposed change does not result in a reduction in the margin of

safety.

The remaining changes are purely editorial and do not affect the

design or operation of any equipment or alter the technical

requirements of any TS. As a result, these proposed changes do not

result in a reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

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

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

NRC Project Director: Gail H. Marcus

Illinois Power Company and Soyland Power Cooperative, Inc., Docket

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

Illinois

Date of amendment request: October 27, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification 5.2.2.e, ``Unit Staff,'' to revise the

requirements for controls on the working hours of unit staff who

perform safety related functions. The proposal would clarify the

approval requirements for deviations from the overtime guidelines and

eliminate the requirement for a monthly review of individual overtime,

consistent with GL 82-12, ``Nuclear Power Plant Staff Working Hours,''

dated June 15, 1982.

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 change to the plant

design or operation. The proposed changes do not affect the level of

approval required for deviations from the overtime guidelines. As

the Technical Specifications will continue to require deviations

from the guidelines for overtime control to be approved and

documented, the proposed changes do not adversely affect the level

of alertness for the unit staff who perform safety-related

functions. The current requirement for the plant manager (or his

designee) to perform a monthly review of individual overtime is an

after the fact review that has not been proven to provide any

significant benefit with respect to the control of individual

overtime. In addition, the proposed changes do not directly affect

the automatic operation of equipment or systems assumed to mitigate

the consequences of previously evaluated accidents. As a result, the

proposed changes do not affect any of the parameters or conditions

that contribute to initiation of an accident previously evaluated,

and thus, the proposed changes cannot increase the probability or

the consequences of any accident previously evaluated.

(2) The proposed changes do not involve a change to the plant

design or operation. The proposed changes do not affect the level of

approval required for deviations from the overtime guidelines and do

not adversely affect the level of alertness for the unit staff who

perform safety-related functions. As a result, the proposed changes

do not affect any of the parameters or conditions that could

contribute to initiation of an accident, and thus cannot create the

possibility of an accident not previously evaluated.

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

in a margin of safety. As noted previously, the proposed changes do

not change the level of approval required for deviations from the

overtime guidelines. Only the requirement for an after-the-fact

monthly review is proposed to be deleted. To the extent that

personnel alertness may be regarded as a margin of safety, deleting

this requirement will not result in a significant reduction in a

margin of safety since overtime controls consistent with the

guidelines and requirements of GL 82-12 will continue to remain in

place.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

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

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

NRC Project Director: Gail H. Marcus

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

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

Michigan

Date of amendment requests: May 19, 1995, as supplemented October

20, 1995 (AEP:NRC:1213A)

Description of amendment requests: The proposed amendments would

modify the Technical Specification (TS) action statement associated

with the main steam safety valves (MSSVs). The action statement would

reflect different requirements based on operating mode and the power

range neutron flux high setpoint with inoperable MSSVs would be revised

in response to an issue raised in Westinghouse Nuclear Safety Advisory

Letter 94-001. The supplement also requested the addition of an

exemption to TS 4.0.4 in the surveillance requirements for the MSSVs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

Correction of the setpoint methodology does not represent a

credible accident initiator. The new methodology reduces the

allowable power level setpoints and is conservative compared to the

presently evaluated setpoints. The consequences of any previously

evaluated accident are not adversely affected by this action because

the decrease in the setpoints resulting from the new calculational

methodology will ensure that the MSSVs are capable of relieving the

pressure at the allowable power levels. Based on these

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

significant increase in the probability or consequences of an

accident previously evaluated.

Correcting the overly restrictive action statements of T/S 3.7.1

does not involve a significant increase in the probability of an

accident. The proposed changes modify existing text to more

accurately reflect the intention of the restrictions imposed by the

action statements. The changes do not create any situation that

would initiate a credible accident sequence.

The proposed 4.0.4 exemption is necessary to make the T/Ss

accurately reflect limitations associated with conduct of the

surveillance in Mode 3. Additionally, the change is needed to

address the fact that unscheduled outages can and do occur and, when

they do, surveillances can expire with no way to correct the

situation until the unit returns to power. Since the purpose of the

4.0.4 exemption is to allow surveillances to be conducted after an

extended period of reactor shutdown, the decay heat to be removed by

the MSSVs will be less than (and therefore conservative compared to)

the conditions experienced when the surveillances are already

allowed by the T/Ss. These allowed conditions include conduct of the

surveillance during power operation or immediately after shutdown.

Therefore, we believe that any increase in the probability of

occurrence or consequences of an accident previously analyzed would

be insignificant.

Criterion 2

The change in Table 3.7-1 reduces the allowable power levels

that can be achieved in the event that one or more main steam safety

valve(s) is inoperable. This change is a result of vendor guidance

to correct an error in the existing methodology used to determine

the setpoints for the power level.

[[Page 65682]]

Changing the methodology used to determine the setpoints, and lowering

the setpoints themselves, do not create a new condition that could

lead to a credible accident. Therefore, it is concluded that the

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

kind of accident from any accident previously evaluated.

The action statements remain in effect to perform the intended

function of protecting the plant's secondary side when the main

steam safety valves are inoperable. They have only been modified to

correct the overly restrictive language that specifies when, in each

mode, specific actions must be taken. Therefore, the proposed change

does not create a new or different type of accident.

Because the proposed 4.0.4 exemption requires neither physical

changes to the plant nor changes to the safety analyses, we believe

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

of accident from any previously evaluated.

Criterion 3

The margin of safety presently provided is not reduced by the

proposed change in the setpoints. The change will correct the

limiting power levels that are to be implemented when MSSVs are

inoperable. This action does not adversely affect the margin that

was previously allocated for the ability of the MSSVs to relieve

secondary side pressure. Based on these considerations, it is

concluded that the changes do not involve a significant reduction in

a margin of safety.

The margin of safety is also not significantly reduced by the

proposed change to the action statements of the T/S. The proposed

revision clarifies when specific actions are to be taken in response

to inoperable main steam safety valves. The changes do not decrease

the effectiveness of the actions to be taken; therefore, they do not

significantly reduce any margin of safety.

The margin of safety is not adversely affected by the proposed

exemption to T/S 4.0.4, since the surveillance conditions allowed by

the exemption are bounded by the normal surveillance conditions seen

immediately after shutdown or during power operation.

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

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

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

amendment requests involve no significant hazards consideration. The

initial application was noticed in the Federal Register on June 21,

1995 (60 FR 32368).

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Brian E. Holian, Acting

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

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

Michigan

Date of amendment requests: November 10, 1995 (AEP:NRC:0896X)

(Supersedes application dated June 15, 1995.)

Description of amendment requests: The proposed amendments would

change the 18-month emergency diesel generator (EDG) surveillance test

from a 24-hour run to an 8-hour run and would add voltage and frequency

measurement and power factor monitoring.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The safety function of the EDGs is to supply AC electrical power

to plant safety systems whenever the preferred AC power supply is

unavailable. Through surveillance requirements, the ability of the

EDGs to meet their load and timing requirements is tested and the

quality of the fuel and the availability of the fuel supply are

monitored. Reduction of the 24 hour run to 8 hours will not reduce

the surveillance effectiveness and will sufficiently exercise the

EDG and its support systems to identify potential conditions that

could lead to performance degradation (See Attachment 4 [of

amendment request]). Further, monthly full-load testing will provide

confidence in diesel reliability and performance capability. Based

on these considerations, it is concluded that the proposed changes

do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Criterion 2

The proposed changes do not involve physical changes to the

plant or changes in plant operating configuration. The changes only

involve EDG surveillance test requirements. These changes will not

affect EDG operability and are designed to improve surveillance

effectiveness. Also, paralleling the diesel to the system grid

during normal operations has been performed to fulfill monthly

surveillance requirements when the resistive load banks were not

available.

It is recognized that, during the 1 hour monthly surveillance

test period, the diesel could be exposed to electrical system

transients (e.g., transients induced by inclement weather

conditions) which could cause the paralleled diesel output breaker

to trip open. Such a scenario, although unlikely, is mitigated by

the availability of the alternate EDG which is placed in the auto

start mode prior to the surveillance. In addition, during testing,

an operator is continuously monitoring the diesel control panel and

can, if necessary, reset the affected EDG lockout relays to restore

EDG availability. Therefore, it is concluded that the proposed

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

accident from any accident previously evaluated.

Criterion 3

Although the duration of the EDG 18 month 24 hour surveillance

test would be reduced, the EDG components will continue to be

sufficiently exercised such that the ability to detect incipient and

degraded conditions will be maintained (See Attachment 4, Figure 2

[of amendment request]). Also, the added review of diesel reactive

loading ensures that test conditions closely match potential

emergency conditions. In addition, the monthly full-load testing

will provide confidence in diesel reliability and performance

capability without impacting diesel operability. During the monthly

test, the impact on plant safety due to potential exposure to

transient grid conditions is considered to be insignificant based on

the likelihood of such transients coincident with the testing and

the mitigating factors discussed in Criterion 2 above.

Based on the above considerations, it is concluded that the

proposed changes do not involve a significant reduction in a margin

of safety.

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

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

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

amendment requests involve no significant hazards consideration. This

notice supersedes the staff's notice published in the Federal Register

on July 19, 1995 (60 FR 37096).

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Brian E. Holian, Acting

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New London County, Connecticut

Date of amendment request: October 25, 1995

Description of amendment request: The amendment request would

revise the Technical Specifications (TS) to relocate the flow-biased

average power range monitor (APRM) scram and rod block setpoint

requirements for reactor operation with excessive core peaking, which

will also include surveillance requirements to verify the setpoints.

The amendment would also delete TS Figure 2.1.2, and any references to

the figure. APRM meter setting adjustments would be changed to allow

setpoint adjustment to be made at power levels less than or equal to

90% of the rated, and the

[[Page 65683]]

requirement that the scram setting adjustment be A.C. Sources,'' ``D.C. Sources,'' and ``Onsite

Distribution

[[Page 65690]]

Systems,'' respectively) are being revised to provide guidance on how

and why degraded offsite power voltage and the number of startup

transformers in service affect compliance to GDC 17 and to give the

basis for the additional action statements.

Date of issuance: November 28, 1995

Effective date: November 28, 1995

Amendment Nos.: Unit 1 - Amendment No. 102; Unit 2 - Amendment No.

90; Unit 3 - Amendment No. 73

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

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 2, 1995 (60 FR

39431) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 28, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location:Phoenix Public Library, 12 East

McDowell Road, Phoenix, Arizona 85004

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

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

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

Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: August 30, 1994, as

supplemented August 4, 1995.

Brief description of amendments: This application upgrades the

current custom Technical Specifications (TS) for Dresden and Quad

Cities to the Standard Technical Specifications contained in NUREG-

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

4.'' This application upgrades only Section 3/4.2,

``Instrumentation.''Date of issuance: November 20, 1995

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

30, 1996.

Amendment Nos.: 142, 136, 164, and 160

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

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 30, 1995 (60 FR

45177) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 20, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: for Dresden, Morris Area

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

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

Illinois 61021.

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

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

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

and 2, Rock Island County, Illinois

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

supplemented July 20, 1995.

Brief description of amendments: This application upgrades the

current custom Technical Specifications (TS) for Dresden and Quad

Cities to the Standard Technical Specifications (STS) contained in

NUREG-0123, ``Standard Technical Specification General Electric Plants

BWR/4.'' This application upgrades only Section 3/4.7, ``Containment

Systems.''

Date of issuance: November 27, 1995

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

30, 1996, for Dresden Station and June 30, 1996, for Quad Cities

Station.

Amendment Nos.: 143, 137, 165, 161

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

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 2, 1995 (60 FR

39433) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 27, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: for Dresden, Morris Area

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

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

Illinois 61021.

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

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

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

supplemented by letter

dated June 29, 1995

Brief description of amendments: The amendments would revise and

clarify portions of Technical Specification Section 6.0,

``Administrative Controls.''

Date of Issuance: December 1, 1995

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

30 days

Amendment Nos.: 211, 211, and 208

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 15, 1995 (60 FR

14020) The June 29, 1995, letter provided clarifying information that

did not change the scope of the January 12, 1995, application and the

initial proposed no signficant hazards consideration determination. The

Commission's related evaluation of the amendments is contained in a

Safety Evaluation dated December 1, 1995No significant hazards

consideration comments received: No

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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

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

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

supplemented by letter dated November 15, 1995

Brief description of amendments: The amendments revise Technical

Specification (TS) 6.9.2 to include references to updated or recently

approved mathodologies used to calculate cycle-specific limits

contained in the Core Operating Limits Report. The subject references

have previously been reviewed and approved by the NRC staff.

Date of Issuance: December 4, 1995

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

30 days

Amendment Nos.: 212, 212, 209

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 11, 1995 (60 FR

52928) The November 15, 1995, letter provided clarifying information

that did not change the scope of the September 1, 1995, application and

the initial proposed no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 4, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location:Oconee County Library, 501 West

South Broad Street, Walhalla, South Carolina 29691

[[Page 65691]]

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: July 26, 1995, as supplemented

by letter dated October 4, 1995

Brief description of amendments: These amendments concern revising

certain surveillance intervals and allowable outage times for the RPS

and ESFAS equipment.

Date of issuance: November 29, 1995

Effective date: November 29, 1995Amendment Nos. 179 and 173Facility

Operating Licenses Nos. DPR-31 and DPR-41: Amendments revised the

Technical Specifications.

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

54720) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 29, 1995 No significant

hazards consideration comments received: No

Local Public Document Room location:Florida International

University, University Park, Miami, Florida 33199.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile

Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: May 24, 1995, as supplemented

July 24, 1995

Brief description of amendment: The amendment revises the Technical

Specifications to extend the test interval for the source range neutron

flux instrumentation from 7 days prior to startup to 6 months prior to

startup.

Date of Issuance: November 24, 1995

Effective date: As of its date of issuance, to be implemented

within 30 days.

Amendment No.: 199

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 21, 1995 (60 FR

32365) The July 24, 1995, letter provided clarifying information that

did not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of

this amendment is contained in a Safety Evaluation dated November 24,

1995.No significant hazards consideration comments received: No.

Local Public Document Room location: Law/Government Publications

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

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: June 14, 1993, supplemented April 12,

1994

Brief description of amendment: The amendment revised the technical

specifications (TSs) to include wording consistent with 10 CFR Part 20,

and to deleted TSs governing miscellaneous radioactive material sealed

sources.

Date of issuance: November 28, 1995

Effective date: November 28, 1995

Amendment No.: 174

Facility Operating License No. DPR-46. Amendment revised the

Technical Specifications.

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

FR 46237) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 28, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location:Auburn Public Library, 118 15th

Street, Auburn, NE 68305.

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

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

Date of amendment request: July 24, 1995, as supplemented by letter

dated October 30, 1995.

Description of amendment request: The amendment revises the

Appendix A Technical Specifications (TS) relating to reactor coolant

system leakage. Specifically, the amendment deletes Table 3.4-1,

``Reactor Coolant System Pressure Isolation Valves'' from the Seabrook

Station, Unit No. 1 TS section 3.4.6.2. Also, reference to Table 3.4-1

is deleted from Limiting Condition for Operation 3.4.6.2 f and from

Surveillance Requirement 4.4.6.2.2. The information contained in Table

3.4-1 is to be relocated to the Technical Requirements Manual.

Additionally, a footnote providing certain exceptions from the

requirements of SR 4.4.6.2.2d for the RHR Pump A and RHR Pump B Suction

Isolation Valves previously located on Table 3.4-1 is relocated as a

footnote to SR 4.4.6.2.2d.

Date of issuance: November 28, 1995

Effective date: As of its date of issuance, to be implemented

within 60 days.

Amendment No.: 44

Facility Operating License No. NPF-86. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 30, 1995 (60 FR

45180). The licensee's letter dated October 30, 1995, provided a minor

revision to the application that was within the scope of the original

notice and did not change the initial proposed no significant hazards

consideration determination. The October 30, 1995, letter also

contained a request for an additional change that will be addressed

separately.The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 28, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location:Exeter Public Library, Founders

Park, Exeter, NH 03833.

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

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

Date of amendment request: September 20, 1995

Description of amendment request: The amendment modifies the

Appendix A Technical Specifications for the Engineered Safety Features

Actuation System Instrumentation. Specifically, the amendment revises

the Seabrook Station Technical Specifications to relocate Functional

Unit 6.b, ``Feedwater Isolation - Low RCS Tavg Coincident with a

Reactor Trip'' from Technical Specification 3.3.2. ``Engineered Safety

Features Actuation System Instrumentation'' to the Technical

Requirements Manual which is a licensee controlled document.

Date of issuance: November 29, 1995

Effective date: As of its date of issuance, to be implemented

within 60 days.

Amendment No.: 45

Facility Operating License No. NPF-86. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 24, 1995 (60 FR

54524). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 29, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location:Exeter Public Library, Founders

Park, Exeter, NH 03833.

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

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

Date of amendment request: June 7, 1995.

Description of amendment request: The amendment increases the

temperature limit, as specified by the

[[Page 65692]]

footnotes to Technical Specification Surveillance Requirement 4.4.7 and

to Table 3.4-2, above which reactor coolant sampling and analysis for

dissolved oxygen is required and dissolved oxygen limits apply.

Date of issuance: November 29, 1995

Effective date: November 29, 1995

Amendment No.: 46

Facility Operating License No. NPF-86. Amendment revised the

Technical Specifications.

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

37098). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 29, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location:Exeter Public Library, Founders

Park, Exeter, NH 03833.

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

MillstoneNuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of application for amendment: June 8, 1995

Brief description of amendment: The amendment modifies Surveillance

Requirement (SR) 4.5.1.c and deletes Technical Specification (TS) 3/

4.8.4.3, ``AC Circuits Inside Containment.'' The changes to SR 4.5.1.c

clarify the requirements for securing the safety injection accumulator

isolation valve breakers (3SIL*MV8808A, B, C, and D) in the tripped

position for the applicable modes. The amendment also deletes TS 3/

4.8.4.3 since reasonable assurance is provided to protect the

electrical penetrations and penetration conductors against an

overcurrent condition and single failure of a circuit breaker.

Date of issuance: November 29, 1995

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

within 60 days.

Amendment No.: 121

Facility Operating License No. NPF-49. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 2, 1995 (60 FR

39444) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 29, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location:Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360.

Philadelphia Electric Company, Docket No. 50-353, Limerick

Generating Station, Unit 2, Montgomery County, Pennsylvania

Date of application for amendment: June 23, 1995

Brief description of amendment: This amendment involves a one-time

change affecting the Allowed Outage Time (AOT) for the Emergency

Service Water (ESW) system, Residual Heat Removal Service Water (RHRSW)

System, the Suppression Pool Cooling, the Suppression Pool Spray, and

Low Pressure Coolant Injection modes of the Residual Heat Removal

System, and Core Spray System to be extended from 3 and 7 days to 14

days during the Unit 2 refueling outage scheduled to begin in January

1996. This proposed extended AOT allows adequate time to install

isolation valves and cross-ties on the ESW and RHRSW Systems to

facilitate future inspections or maintenance.

Date of issuance: November 30, 1995

Effective date: November 30, 1995

Amendment No. 70

Facility Operating License No. NPF-85. This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 2, 1995 (60 FR

39448) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 30, 1995No significant

hazards consideration comments received: No

Local Public Document Room location:Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

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

Date of application for amendment: March 3, 1995, as supplemented

April 12, 1995, and November 20, 1995.

Brief description of amendment: The amendment revises the TS to

extend the calibration frequency for the following:

(1) Containment water level monitor instrumentation (specified in

TS Table 4.1-1)

(2) Containment building ambient temperature sensors (specified in

TS Table 4.1-1)

(3) Seismic monitoring instrumentation (specified in TS Table 4.10-

2)

In addition, the amendment added a new surveillance requirement to

TS Table 4.1-1 for testing the core exit thermocouples.

These changes allow operation on a 24-month fuel cycle and follow

the guidance provided in Generic letter 91-04, ``Changes in Technical

Specification Surveillance Intervals to Accommodate a 24-Month Fuel

Cycle,'' as applicable.

Date of issuance: December 1, 1995

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

30 days.

Amendment No.: 164

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

Technical Specifications.

Date of initial notice in Federal Register: May 10, 1995 (60 FR

24917) The April 12 and November 20, 1995, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated December 1, 1995.No

significant hazards consideration comments received: No

Local Public Document Room location:White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: July 27, 1995

Brief description of amendment: The amendment changes the Technical

Specifications to incorporate updated pressure vs. temperature

operating limit curves.

Date of issuance: November 28, 1995

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

60 days

Amendment No.: 88

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

Technical Specifications.

Date of initial notice in Federal Register: September 13, 1995 (60

FR 47624) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 28, 1995. No

significant hazards consideration comments received: No

Local Public Document Room location:Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

Generating Station, Salem County, New Jersey

Date of application for amendment: March 31, 1994, supplemented by

letters dated August 29, and October 16, 1995.

Brief description of amendment: This amendment changes Technical

Specification (TS) 3.5.1, ``ECCS -

[[Page 65693]]

Operating,'' and associated Bases, to establish a new allowed out-of-

service time. Action c.2 for TS 3.5.1 allows any one Low Pressure

Coolant Injection subsystem, or one Core Spray subsystem, to be

inoperable in addition to an inoperable High Pressure Coolant Injection

system, for 72 hours.

Date of issuance: November 30, 1995

Effective date: As of the date of issuance and shall be implemented

within 60 days.

Amendment No.: 89

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

Technical Specifications.

Date of initial notice in Federal Register: June 8, 1994 (59 FR

29631).The supplemental letters did not change 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 November 30, 1995.No significant hazards consideration comments

received: No

Local Public Document Room location:Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

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

Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: March 24, June 9, and June 30,

1995

Brief description of amendment: The amendment revised the Technical

Specifications to allow a one-time extension for the performance of

certain Surveillance Requirements (SRs). Affected SRs include

penetration leak rate testing, valve operability testing, instrument

calibration, response time testing, and logic system functional tests.

The proposed changes are to support refueling outage 5 scheduled to

begin no later than February 15, 1996.

Date of issuance: November 29, 1995

Effective date: November 29, 1995

Amendment No. 75

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

Technical Specifications.

Date of initial notice in Federal Register: May 10, 1995 (60 FR

24919) and August 16, 1995 (60 FR 42612)The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

November 29, 1995. No significant hazards consideration comments

received: No

Local Public Document Room location:Perry Public Library, 3753 Main

Street, Perry, Ohio 44081

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

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

Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: October 21, 1994

Brief description of amendment: The amendment revised Technical

Specification 3/4.6.1.2, ``Primary Containment Leakage,'' and its

associated Bases to reflect the partial exemptions to the requirements

of 10 CFR Part 50, Appendix J, Sections III.A.5(b)(2), III.B.3,

III.C.3, III.A.1(d), III.D.1(a), and III.D.3 that were granted by the

NRC on December 4, 1995.

Date of issuance: December 8, 1995

Effective date: ]December 8, 1995

Amendment No.: 76

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

Technical Specifications.

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

42611) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 8, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

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

Date of application for amendment: July 28, 1995

Brief description of amendment: The amendment clarifies the

limiting condition for operation for TS 3.8.1.1 and 3.8.1.2 from

``independent'' circuit to ``qualified'' circuit; explains in the Bases

the requirements for operability of an offsite circuit; deletes the

STAGGERED TEST BASIS scheduling requirement to perform emergency diesel

generatorsurveillances; explains in the Bases an acceptable method for

verification of Emergency Diesel Generator speed for surveillance

requirements (SR) 4.8.1.1.2.a.4 and 4.8.1.1.2.c.4; removes a

surveillance test extension that has expired for SR 4.8.1.1.1.b; adds

an exception for SR 4.8.1.1.2.c.5 and 4.8.1.1.2.c.7 to SR 4.8.1.2; and

revises Bases 3.0.5 to reflect the clarification from ``independent''

circuit to ``qualified'' circuit.

Date of issuance: December 8, 1995

Effective date: December 8, 1995

Amendment No.: 203

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

Technical Specifications.

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

56370) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 8, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location:University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606.

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: October 2, 1995

Brief description of amendment: The amendment revises Technical

Specification (TS) Section 5.0, ``Design Features,'' by adding a site

location description, removing site area maps, removing containment and

reactor coolant system design parameters, removing the description of

the meteorological tower location, removing component cyclic or

transient limits, and revising the fuel assembly description to include

the use of ZIRLO clad fuel rods.

Date of issuance: December 8, 1995

Effective date: December 8, 1995

Amendment No.: 204

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

Technical Specifications.

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

56371) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 8, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location:University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment: June 23, 1995

Brief description of amendment: The amendment revises Technical

[[Page 65694]]

Specification (TS) 4.1.3.1.2, 4.4.6.2.2.b, 4.4.3.2, 4.6.2.1.d, 4.6.4.2,

and Table 4.3-3 in accordance with guidance provided in NRC Generic

Letter (GL) 93-05, ``Line Item Technical Specification Improvements to

Reduce Surveillance Requirements for Testing During Power Operations.''

Additionally, the amendment revises TS 4.1.1.1.1, 4.1.1.2, 3/4.1.3.1

and the associated Bases to implement portions of NUREG-1431,

``Standard Technical Specifications - Westinghouse Plants.''

Date of issuance: December 7, 1995

Effective date: December 7, 1995

Amendment No.: 105

Facility Operating License No. NPF-30. The amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: August 30, 1995 (60 FR

45187). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 7, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location:Callaway County Public Library,

710 Court Street, Fulton, Missouri 65251.

Dated at Rockville, Maryland, this 13th day of December 1995.For

the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II, Office of Nuclear

Reactor Regulation

[Doc. 95-30755 Filed 12-19-95; 8:45 am]

BILLING CODE 7590-01-F