Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating

LicensesInvolving No Significant Hazards Considerations

I. Background

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

Commission (the Commission or NRC staff) is

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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 October 14, 1995, through October 27, 1995.

The last biweekly notice was published on Wednesday, October 25, 1995

(60 FR 54714).

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 December 8, 1995, 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 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

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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 50-529 and

STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 2 and

3, Maricopa County, Arizona

Date of amendments request: October 3, 1995

Description of amendments request: The amendment would delete the

provisions relating to certain previous sale and leaseback transactions

that were by added by Amendment No. 3 for NPF-51 and Amendment No. 1

for NPF-74.

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.

This amendment request does not involve a significant increase

in the probability or consequences of an accident previously

evaluated because the proposed change is administrative in nature.

The proposed change deletes Sections 2.B.(7)(a) and (b) of License

No. NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74.

These sections describe the structure of the financing of El Paso's

interest in Palo Verde, specifically authorizing sale and leaseback

transactions. The proposed change does not affect the assumptions

used in the accident

analyses, nor does the proposed change result in changes to the

physical configuration of the facility, design parameters, technical

specifications, or operation and maintenance of the facility.

Therefore, the amendment request does 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.

This amendment request does not create the possibility of a new

or different kind of accident from any accident previously analyzed

because the proposed change is administrative in nature. The

proposed change deletes Sections 2.B.(7)(a) and (b) of License No.

NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These

sections describe the structure of the financing of El Paso's

interest in Palo Verde Units 2 and 3, specifically authorizing sale

and leaseback transitions. The proposed change does not involve

modifications to any of the existing equipment nor does the change

affect operation or maintenance of the facility. Therefore, the

amendment request does not create the possibility of a new or

different kind of accident not previously analyzed.

3. The proposed change does not involve a significant reduction

in a margin of safety.

This amendment request does not involve a significant reduction

in a margin of safety because it is administrative in nature. The

proposed change deletes Sections 2.B.(7)(a) and (b) of License No.

NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These

sections describe the structure of the financing of El Paso's

interest in Palo Verde, specifically authorizing the sale and

leaseback transactions. The proposed change does not involve changes

to any existing plant equipment or accident analyses that provide

for or establish margins of safety. There is no change to the

operation or maintenance of the facility and the existing margins of

safety are not changed by the proposed change. Therefore, the

amendment request does 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

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert

Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of amendment request: October 2, 1995

Description of amendment request: The proposed amendment would

revise the Calvert Cliffs Nuclear Power Plant, Unit No. 2, Technical

Specifications on a one-time basis by increasing the 7 day allowed

outage time (AOT) of the control room emergency ventilation system

(CREVS) to an AOT of 30 days. This requested one-time increase in the

AOT is applicable only for the loss of the emergency power supply to

one train of the CREVS during the Unit No. 1 spring 1996 refueling

outage.

The requested extension in the AOT is necessary to allow the

licensee to perform modifications to the electrical distribution system

during the upcoming Unit 1 refueling outage while Unit No. 2 continues

to operate. The modifications include connecting a fourth safety-

related (SR) emergency diesel generator (EDG) to engineered safety

features (ESF) Bus No. 11. The work related to this effort will require

that the bus be deenergized for several days isolating it from its

normal and

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emergency EDG power supplies. One train of the CREVS is connected to

ESF Bus No. 11 and will not have its power supplies available for a

period of time. The normal (offsite) power is expected to be restored

in about 3 days, but the emergency power (onsite EDG) may take up to 30

days.

The licensee is taking additional actions to assure the

availability of the normal offsite power source and is also adding a

nonsafety-related (NSR) EDG as an alternate onsite power source during

the period that the SR EDG is not available. The licensee expects that

the tie-in of the NSR EDG will take about 8 days. Thus, even if the

normal offsite power source is lost, the temporary onsite NSR EDG will

be available to provide power to the affected train of the CREVS.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The Control Room Emergency Ventilation System (CREVS) is used to

mitigate the consequences of an accident. It is designed so that the

Control Room remains habitable for operators and to maintain the

environment needed for continued equipment operation. The system is

redundant (two 100% capacity trains) and is powered from both normal

(offsite) and emergency (emergency diesel generators) power sources.

We [the licensee] are proposing an amendment which would allow the

emergency power to be removed from one of the redundant CREVS for an

additional 23 days (beyond the 7 days allowed by the Technical

Specifications). Other than the removal of the emergency electrical

power source, we are not affecting or modifying the operation of the

CREVS. The CREVS is not an accident initiator for any previously

evaluated accident. Therefore, the proposed change does not involve

an increase in the probability of an accident previously evaluated.

The CREVS is designed to mitigate the consequences of design

basis accidents. For that purpose, redundant trains are provided to

protect against a single failure. During the Technical Specification

seven day Allowed Outage Time (AOT), an operating unit is allowed by

the Technical Specifications to remove one of the CREVS trains from

service, thereby eliminating this single failure protection. The

consequences of a design basis accident coincident with a failure of

the redundant CREVS train during the additional 23-day period are

the same as those during the 7-day AOT. Therefore, the proposed

change does not significantly increase the consequences of an

accident previously evaluated.

Therefore, the proposed change does not increase the probability

or consequences of an accident previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

The CREVS is not being modified by this proposed change nor will

any unusual operator actions be required. The system will continue

to operate in the same manner. The CREVS is not an initiator to any

accident, but is designed to respond should an accident occur.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

safety.

The operability of the CREVS during Modes 1 through 4 ensures

that the Control Room will remain habitable for operators and to

maintain the environment needed for continued equipment operation

under all plant conditions. The proposed change does not affect the

function of the CREVS. During the period of the Technical

Specifications AOT when one CREVS train is inoperable, the margin of

safety is reduced. This time period is a temporary relaxation of the

single failure criteria, which, consistent with overall system

reliability considerations, provides a limited time to maintain or

repair the equipment and conduct testing. We are requesting an

extension of this limited time. The proposed change will allow one

train of the CREVS to be without an emergency power supply for an

additional 23 days beyond the 7-day AOT (total of 30 days). This

train of CREVS will be functional and will have the normal power

supply available for all but approximately three days to allow work

and necessary testing on the bus. The other train of the CREVS will

have both its normal and emergency power supplies during this

period.

To provide additional assurance that all reasonable steps have

been taken to prevent the loss of the normal power supply to the

CREVS, we will restrict maintenance activities on three of the four

offsite transmission lines. This restriction will cover the period

we are in the Action Statement for the CREVS (Action Statement

3.7.6.1.a and b). To provide an alternative power source during the

majority of this period, we will connect the Alternate AC power

source (No. 0C Diesel Generator) to ESF Bus No. 11 and confirm its

availability as soon as possible after the work on ESF Bus No. 11

begins (we [the licensee] expect that to take about eight days).

This power source is independent from the offsite power supplies. In

addition, we will restrict planned maintenance on the No. 12 CREVS

during the period we are in the Action Statement to ensure that the

No. 12 CREVS is not removed from service.

We believe that the reduction in the margin of safety

represented by this one-time extension of the AOT is not significant

based on our management of plant risk, the reliability of the normal

CREVS power supply, the availability of the redundant CREVS with

both its normal and emergency power, and the mitigating features

described above. Therefore, the proposed change does not involve a

significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Calvert County Library, Prince

Frederick, Maryland 20678

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

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

NRC Project Director: Ledyard B. Marsh

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

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendments request: October 23, 1995

Description of amendments request: The amendments would delete the

applicability of the primary coolant water chemistry limits when the

primary system is being chemically decontaminated and the reactor

vessel is defueled.

Basis for proposed no significant hazards consideration

determination: As required by 10CFR 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 will allow the reactor coolant system

conductivity and chlorides to exceed the limits specified in

Technical Specification Table 3.4.4-1 in support of performing

chemical decontamination activities. The reactor coolant system

water chemistry limits have been established to prevent long-term

damage to the reactor coolant system materials that are in contact

with the coolant. Upon concluding the chemical decontamination

activities, reactor coolant system conductivity and chloride values

would be restored to within the limits specified in Technical

Specification Table 3.4.4-1. Existing regulatory requirements,

specifically a review in accordance with 10 CFR 50.59 to determine

whether an activity involves an unreviewed safety question, provide

adequate assurance that solvents selected for use in a chemical

decontamination activity will not degrade the structural integrity

of the reactor coolant system. Therefore, since the structural

integrity of the reactor coolant system will not be adversely

impacted by the chemical decontamination activities, the proposed

amendments do not involve a significant increase in the probability

of an accident previously evaluated.

As discussed above, the reactor coolant system water chemistry

limits have been

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established to prevent long-term damage to the reactor coolant system

materials that are in contact with the coolant. The solvents being

used for a chemical decontamination activity are selected to ensure

their effectiveness and to ensure that damage will not occur to the

structural materials comprising the reactor coolant pressure

boundary. As such, the operation of safety equipment used to

mitigate a design basis accident or transient will not be affected

by the proposed change of the reactor coolant system water chemistry

limits during performance of chemical decontamination activities.

Therefore, the proposed revision to the reactor coolant system

chemistry limits will not involve a significant increase in the

consequences of an accident previously evaluated.

2. The proposed change will allow the reactor coolant system

conductivity and chlorides to exceed the limits specified in

Technical Specification Table 3.4.4-1 in order to perform chemical

decontamination activities. The reactor coolant system water

chemistry limits have been established to prevent long-term damage

to the reactor coolant system materials that are in contact with the

coolant. Even though the solvents used for chemical decontaminations

may result in reactor coolant system conductivity and chloride

measurement values in excess of the limits specified in the

Technical Specifications, the existing regulatory requirements of 10

CFR 50.59 will continue to ensure that solvents being used for

performing chemical decontamination have been properly evaluated and

that these solvents do not adversely affect the material properties

or structural integrity of the reactor coolant system. Therefore,

the proposed amendments revising the reactor coolant system water

chemistry limits during performance of chemical decontamination

activities will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3. The reactor coolant system water chemistry limits have been

established to prevent long-term damage to the reactor coolant

system materials that are in contact with the coolant. The solvents

used for chemical decontaminations result in reactor coolant system

conductivity and chloride measurement values in excess of the limits

specified in the Technical Specifications; however, the solvents

being used of performing chemical decontamination have been properly

evaluated to ensure they will not significantly affect the material

properties of the reactor coolant system piping (i.e., corrosion)

nor will they significantly affect the structural integrity (i.e.,

wall thinning) of the reactor coolant system piping. Therefore, the

proposed license amendments do not involve a significant reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297

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

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

NRC Project Director: David B. Matthews

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

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

Date of amendment request: November 2, 1994, as supplemented on

January 4, 1995

Description of amendment request: The amendment would revise the

Technical Specifications (TSs) to make editorial changes, delete

portions of the TSs that have become unnecessary due to previously

approved amendments, change managerial titles, update references and

reporting requirements, revise the Station Nuclear Safety Committee

(SNSC) composition to specify disciplines rather than specific job

titles, modify the record keeping requirements of the Nuclear

Facilities Safety Committee, implement changes referenced in Generic

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

Administrative Control Requirements for Emergency and Security Plans,''

and to correct the shift manning requirements table.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. There is no significant increase in the probability or

consequences of an accident previously evaluated.

The proposed amendments are administrative in nature. They

involve making editorial changes, deleting portions of the Technical

Specifications that have become unnecessary due to previously

approved amendments, changing managerial titles, updating references

and reporting requirements, revising the SNSC composition to specify

disciplines rather than specific job titles, implementing changes

referenced in Generic Letter 93-07, and revising shift manning to

conform with the requirements of 10 CFR 50.54. These changes do not

affect possible initiating events for accidents previously evaluated

or alter the configuration or operation of the facility. The

Limiting Safety Systems Settings and Safety Limits specified in the

current Technical Specifications remain unchanged. Therefore, the

proposed changes to the subject Technical Specifications would not

increase the probability or consequences of an accident previously

evaluated.

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

any accident previously evaluated has not been created.

As stated above, the proposed changes are administrative in

nature. The safety analysis of the facility remains complete and

accurate. There are no physical changes to the facility and the

plant conditions for which the design basis accidents have been

evaluated are still valid. The operating procedures and emergency

procedures are unaffected. Consequently, no new failure modes are

introduced as a result of the proposed changes. Therefore, the

proposed changes would not initiate any new or different kind of

accident.

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

safety.

The proposed changes are administrative in nature. Since there

are no changes to the physical design or operation of the facility,

the Updated Final Safety Analysis Report (UFSAR) design basis,

accident assumptions, or Technical Specification Bases are not

affected. Therefore, the proposed changes would 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610

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

New York, New York 10003

NRC Project Director: Ledyard B. Marsh

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

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

Date of amendment request: August 29, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification Sections 3.1.F and 4.13 to provide for

appropriate inservice inspection for any steam generator tubes

containing sleeves and to provide for reduced allowable primary-to-

secondary leakage rates for steam generators containing sleeves. The

proposed changes are in response to commitments made by Consolidated

Edison by letter dated April 5, 1995, during the review of an amendment

which permitted the use of laser welded steam generator tube sleeves as

a method of tube repair.

Basis for proposed no significant hazards consideration

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

its analysis of the

[[Page 56366]]

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.

Technical Specification Amendment No. 183 allowed sleeving as an

acceptable alternate tube repair method for Indian Point Unit No. 2.

The steam generator sleeve approved for installation is the

Westinghouse process (laser welded sleeve). The sleeve configuration

was designed and analyzed in accordance with the criteria of

Regulatory Guide (RG) 1.121 and the design requirements of Section

III of the American Society of Mechanical Engineers (ASME) Code.

Fatigue and stress analyses of the sleeved tube assembly produced

acceptable results as documented in the Westinghouse topical report

submitted in the original sleeving package. Mechanical testing has

shown that the structural strength of the sleeves under normal,

faulted, and upset conditions is within acceptable limits. Leakage

rate testing for the tube sleeves has demonstrated that primary-to-

secondary leakage is not expected during all plant conditions.

Any leakage through the sleeved region of the tube is fully

bounded by the leak-before-break considerations and, ultimately, the

existing steam generator tube rupture analysis included in the

Updated Final Safety Analysis Report (UFSAR).

The reduction in TS leakage rate requirements from 0.3 gpm

[gallons per minute] (432 gpd [gallons per day]) allowable per SG to

150 gpd per steam generator containing sleeves further ensures that

SG tube integrity is maintained in the event of a main steam line

break (MSLB) or under Loss Of Coolant Accident (LOCA) conditions.

The RG 1.121 criteria for establishing operational leakage rate

limits require a plant shutdown based upon a leak-before-break

consideration to detect a free span crack before a potential tube

rupture. The 150 gpd limit will continue to allow for early leakage

detection and require a plant shutdown in the event of tube leakage

that exceeds the revised Technical Specification limit.

The sleeve sample size has been increased to a minimum of twenty

(20) percent of the inservice sleeves. Increasing the sample size of

the sleeves to be inspected will increase the monitoring of tubes

using sleeves for any further degradation while they remain

inservice. If the sample identifies a sleeve with an imperfection of

greater than 23 percent depth an additional 20 percent of the

sleeves shall be inspected. The sleeves that have identified

imperfections of greater than 23 percent shall be evaluated and

removed from service.

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

or different kind of accident from any accident previously

evaluated.

Implementation of the proposed amendment will not introduce

significant or adverse changes to the plant design basis. The

proposed changes do not involve plant modification or changes to

equipment, and consist of reducing the allowable steam generator

leakage limits for steam generators containing sleeves and defining

the sample size of the steam generator tube sleeve inspection.

The reduction in TS leakage rate requirements from 0.3 gpm (432

gpd) allowable per SG to 150 gpd per SG containing sleeves further

ensures that SG tube integrity is maintained in the event of a MSLB

or under LOCA conditions. The 150 gpd limit is designed to provide

for leakage detection and a plant shutdown in the event of the

concurrence of excessive tube leakage. The limit provides for early

detection and a plant shutdown prior to a postulated defect reaching

critical magnitudes for Main Steam Line Break conditions.

Formalizing the sample size of sleeved tubes inspected during

each scheduled inservice inspection will ensure increased monitoring

of these tubes for any further degradation. The improved monitoring

and evaluation of the tube and the sleeves assures tube structural

integrity is maintained or the tube is removed from service.

With these actions the possibility of a new or different type of

accident from any accident previously evaluated is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Implementation of the proposed changes will not reduce the

margin of safety. This amendment involves the reduction of sleeved

steam generator tube leakage limit and a formalized inservice

inspection program for sleeved tubes. These actions will help ensure

steam generator tube integrity.

Reduction of the leakage rate requirement from 0.3 gpm (432 gpd)

to 150 gallons per day (gpd) per sleeved steam generator will

continue to ensure steam generator tube integrity is maintained in

the event of main steam line break or under LOCA conditions.

Reducing this limit will not result in a reduction in the margin of

safety.

The portions of the installed sleeve assembly which represent

the reactor coolant pressure boundary will be monitored for the

initiation and progression of sleeve/tube wall degradation, thus

satisfying the requirement of Regulatory Guide 1.83. The portion of

the tube bridged by the sleeve joints is effectively removed from

the pressure boundary, and the sleeve then forms the new pressure

boundary. The sleeve enhances the safety of the plant by increasing

the protective boundaries of the steam generator. Keeping the tube

in service with the use of a sleeve, instead of plugging the tube

and removing it from service, increases the heat transfer efficiency

of the steam generator. Monitoring for any increased degradation of

a repaired steam generator tube shall be implemented by sampling

twenty (20) percent of the sleeves inservice. During each scheduled

inservice inspection, any sampled sleeve evaluated and found to have

unacceptable degradation shall be removed from service.

Based on the preceding analysis it is concluded that operation

of Indian Point Unit No. 2 in accordance with the proposed amendment

does not increase the probability of an accident previously

evaluated, does not create the possibility of a new or different

kind of accident from any accident previously evaluated, nor reduce

any margin of plant safety. Therefore, the license amendment does

not involve a Significant Hazards Consideration as defined in 10 CFR

50.92.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610

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

New York, New York 10003

NRC Project Director: Ledyard B. Marsh

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: September 30, 1994, as supplemented by

letter dated September 19, 1995

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) related to the replacement of

the steam generators at McGuire, Units 1 and 2. Currently, the steam

generators in place at the McGuire units are Westinghouse Model ``D''

type preheat steam generators. The tube degradation levels in the

generators has affected the reliability of the units. Therefore, these

generators are scheduled to be replaced with feedring steam generators

designed by Babcock and Wilcox International.

In the licensee's September 19, 1995, supplement, proposed changes

were made to TS Table 2.2-1, ``Reactor Trip System Instrumentation Trip

Setpoints,'' to change the programmed TAVG from 588.2 deg.F to

585.1 deg.F. This temperature was chosen based on returning the

secondary side steam pressure to the original value after replacement

of the steam generators. The licensee stated that 585.1 deg.F was the

assumed value for nominal full power TAVG in all applicable safety

analyses related to replacement of the steam generators.

The licensee also requested that the steam line safety valve lift

settings in Table 3.7-3, which was requested in the September 30, 1994,

application, be withdrawn. The licensee determined that these changes

are no longer needed.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards

[[Page 56367]]

consideration, which is presented below:

Operation of McGuire Nuclear Station in accordance with the

proposed changes to the Technical Specifications will not involve a

significant increase in the probability or consequences of an

accident previously evaluated. The low-low steam generator water

level reactor trip setpoint, the high-high steam generator water

level setpoint for turbine trip and feedwater isolation, and the

low-low steam generator water level setpoint for auxiliary feedwater

initiation are changing to support operation with the replacement

steam generators. These setpoints were chosen both to optimize plant

operation, and ensure that all applicable acceptance criteria are

met for licensing basis safety analysis. These setpoints do not

contribute to the initiation of any accident evaluated in the

McGuire FSAR [Final Safety Analysis Report] and have no adverse

impact on system operation, therefore it can be concluded that these

changes will not significantly increase the probability or

consequences of an accident evaluated in the FSAR.

The reduction in the primary to secondary leakage rate for

McGuire will not increase the probability of an accident evaluated

in the FSAR. This lower limit will require corrective action more

quickly than is currently required in the event that there is a

steam generator tube leak. This change will not significantly affect

the consequences of an accident previously evaluated. The allowable

leakage is being lowered because this leakage has a major impact on

the results of the offsite dose calculation for the locked rotor,

single uncontrolled rod withdrawal, and rod ejection events. The

taller tube bundle in the replacement steam generators will

potentially result in a longer period of tube bundle uncovery during

the above transients. The revised allowable leakages of 0.27 gpm

through all steam generators and 135 gallons per day through any one

generator ensure that the dose analysis results are within the

applicable fraction 10 CFR 100 limits.

The increase in Reactor Coolant System volume due to the

replacement steam generators will not increase the probability or

consequences of an accident previously evaluated. The increase in

volume has no effect on the probability of occurrence of any

accident evaluated in the FSAR. The mass and energy release due to

postulated loss of coolant accidents inside containment has been

analyzed to ensure that the peak containment pressure limit is not

exceeded. All Chapter 15 reanalysis which was required due to the

replacement steam generators assumed the new Reactor Coolant System

volume. Since the results of these analyses show the applicable

acceptance criteria continue to be met, it can be concluded that the

consequences of an accident previously evaluated are not

significantly increased due to this change.

* * * *

Operation of McGuire Nuclear Station in accordance with the

proposed changes to the Technical Specification will not create the

possibility of a new or different accident from any accident

previously evaluated. The proposed changes to revise the low-low

steam generator water level reactor trip setpoint, high-high steam

generator water level setpoint for turbine trip and feedwater

isolation, and low-low steam generator water level setpoint for

auxiliary feedwater initiation ensure that the appropriate

acceptance criteria for FSAR Chapter 15 transients which rely on

these functions are met for operation with the replacement steam

generators. The proposed change to lower primary to secondary

leakage for operation with the replacement steam generators will

require that corrective action be taken more quickly in the event

that steam generator tube leakage is experienced during operation.

As discussed in the technical justification, this will cause the

dose results for transients affected by tube bundle uncovery to be

within acceptable limits. .... The increase in Reactor Coolant

System volume is taken into account in the analysis of the mass and

energy release due to a postulated loss of coolant inside

containment and Chapter 15 events which have been reanalyzed due to

replacement of the steam generators. As discussed above, the

proposed changes will not introduce the possibility of a new or

different accident from any previously evaluated; they will ensure

that transients that take credit for these functions and dose

analyses meet applicable acceptance criteria for operation with the

replacement steam generators.

Operation of McGuire Nuclear Station in accordance with the

proposed changes to the Technical Specifications will not involve a

significant reduction in a margin of safety. The proposed changes

are being made to ensure that transients that rely on low-low steam

generator water level reactor trip setpoint, high-high steam

generator water level setpoint for turbine trip and feedwater

isolation, and low-low steam generator water level setpoint for

auxiliary feedwater actuation meet applicable acceptance criteria.

The reduction in allowable primary to secondary leak rate will

ensure that transients with dose analyses which are affected by the

replacement steam generators meet the current acceptable limits.

.... The proposed change in the Reactor Coolant System volume will

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

increased volume affects the mass and energy release due to a

postulated loss of coolant accident inside containment and the other

Chapter 15 events which were reanalyzed due to replacement of the

steam generators. These events have been analyzed and the results

are within current acceptable limits. As discussed above, the

acceptance criteria for FSAR transients which are affected by these

proposed changes continue to be met, therefore there is no

significant reduction in the margin of safety.

Changes to the steam generator surveillance requirements will

simply delete inspection requirements and repair methods which are

no longer applicable after installation of the replacement steam

generators. The only exception to this is Surveillance Requirement

4.4.5.4.a.9. This requirement is modified to clarify that the

manufacturer will perform the hydrostatic test for the replacement

steam generators. This change will not affect the probability or

consequences of an accident previously evaluated, the purpose of the

preservice inspection is to establish the baseline condition of the

tubing. The baseline condition of the tubing in the replacement

steam generators will be established prior to installation. The

possibility of a new or different accident from any previously

evaluated will not be created. No new accident initiation mechanisms

will be introduced by this change, and the intent of the

requirement, to establish the baseline condition of the tubing, will

be met. Since the baseline condition of the tubing will be obtained

for use in the monitoring of tubing degradation, as is currently

required by the surveillance requirement, there will not be a

significant reduction in the margin of safety.

The changes to Technical Specification 6.9.1.9 are

administrative in nature. These changes are being made to reflect

the most recent revisions of DPC-NE-3002 and DPC-NE-3000, which

include changes associated with the replacement steam generators.

These topical reports revisions will be reviewed and approved for

use regarding Catawba and McGuire Nuclear Stations. Since these

changes are administrative in nature, no significant hazards

considerations are involved.

Proposed revision to TS Table 2.2-1, Reactor Trip System

Instrumentation Trip Setpoints:

proposed change to the Technical Specifications does not

involve a significant increase in the probability or consequences of

an accident previously evaluated. Changing the value for TAVG

in Notes 1 and 2 of Table 2.2-1 will update the value to agree with

the TAVG assumed in the applicable safety analyses for

replacement of the steam generators. Acceptable results were

obtained for all required reanalyses. The probability of an accident

will not be significantly affected by operation with the new

TAVG value, because all equipment will be operated within

acceptable design limits. The consequences of previously evaluated

accidents which are affected by this change have been evaluated, and

have been determined to be within acceptable limits.

This proposed change will not create the possibility of a new or

different kind of accident from any previously evaluated. This

change does not change the physical configuration of the plant, and

all analyses which are affected by replacement of the steam

generators have been determined to have acceptable results assuming

this value for TAVG.

This proposed change to the Technical Specifications will not

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

analyses which were affected by replacement of the steam generators

assumed this value for TAVG and the results were determined to

be within previously acceptable limits.

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.

[[Page 56368]]

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

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas

Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

Date of amendment request: September 4, 1993, as supplemented on

February 16, 1994, and August 4, 1995.

Description of amendment request: The proposed amendment would

revise the Arkansas Nuclear One Industrial Security Plan.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration. The

NRC staff has reviewed the licensee's analysis against the standards of

10 CFR 50.92(c). The NRC staff's review is presented below.

The accident mitigation features of the plant are not affected by

the proposed compensatory measures for protecting the site during

periods when security systems are degraded and therefore no decrease

occurs in the effectiveness of the security program to protect against

radiological sabotage or increased risk to the public health and

safety. This is due to continued compliance with existing regulatory

requirements and other commitments within the security plan. These

changes have no impact on the design basis security threat and

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

accident. New systems, modes of equipment operation, failure modes or

other plan situations are not introduced by these changes. The proposed

changes allow flexibility for the use of compensatory measures and do

not change any safety limits, LCOs, or surveillance requirements on

equipment to operate the plant.

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: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502

NRC Project Director: William D. Beckner

Gulf States Utilities Company, Cajun Electric Power Cooperative,

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

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: October 24, 1995, as supplemented or

supercedes letters dated May 30, and June 20, 1995

Description of amendment request: The proposed amendment would

revise the technical specifications (TSs) on containment systems to

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

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

rate testing program at River Bend Station. The licensee letters dated

May 20, and June 20, 1995, requested an exemption to Appendix J which

subsequently became Option B to the appendix. Those letters were

noticed in the Federal Register on July 5, 1995 (60 FR 35079).

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. This request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change does not involve a change to the plant

design or operation. As a result, the proposed change does not

affect any of the parameters or conditions that contribute to

initiation of any accidents previously evaluated. Thus, the proposed

change cannot increase the probability of any accident previously

evaluated.

The proposed change potentially affects the leak-tight integrity

of the containment structure designed to mitigate the consequences

of a loss-of coolant accident (LOCA). The function of the

containment is to maintain functional integrity during and following

the peak transient pressures and temperatures which result from any

loss-of-coolant accident (LOCA)[LOCA]. The containment is designed

to limit fission product leakage following the design basis LOCA.

Because the proposed change does not alter the plant design, only

the frequency of measuring Type B and C leakage, the proposed change

does not directly result in an increase in containment leakage.

However, decreasing the test frequency can increase the probability

that a large increase in containment leakage could go undetected for

an extended period of time. Based upon the results of the periodic

containment Type A or Integrated Leak Rate Tests (ILRTs) and Type B

and C or Local Leak Rate Tests (LLRTs) surveillance tests, this is

not expected during the remaining life of the plant. The risk

resulting from the proposed changes is as follows:

Type A Testing

NUREG/CR-4330 (NRC86) found that the effect of containment

leakage on overall accident risk is small since risk is dominated by

accident sequences that result in failure or bypass of the

containment. It also determined that on an expected individual dose

basis, the effect of containment leakage is small.

Industry wide, ILRTs have only found a small fraction of the

leaks that exceed current acceptance criteria. Only three percent of

all leaks have a potential for remaining undetected for longer

periods of time. In addition, when leakage has been detected by

ILRTs, the leakage rate has been only about two times the allowable

leakage rate.

NUREG-1493 found that these observations, together with the

insensitivity of reactor accident risk to the containment leakage

rate, show that reducing the Type A leakage test frequency would

have a minimal impact on public risk.

Type B and C Testing

NUREG-1493 found that while Type B and C tests can identify the

vast majority (greater than 95 percent) of all potential leakage

paths, performance-based alternatives to current local leakage-

testing requirements are feasible without significant risk impacts.

The risk model used in NUREG-1493 suggests hat the number of

components tested would be reduced by about 60 percent with less

than a three-fold increase in the incremental risk due to

containment leakage. Since, under existing requirements, leakage

contributes less than 0.1 percent of overall accident risk, the

overall impact is very small. NUREG-1493 found that while the

extended testing intervals for Type B and C tests led to minor

increases in potential offsite [off-site] dose consequences, the

actual increase in on-site (worker) doses exceeded (by at least an

order of magnitude) the potential off-site dose consequences.

EPRI Research Project Report TR-104285, ``Risk Impact Assessment

of Revised Containment Leak Rate Testing Intervals,'' also concluded

that a relaxation of the test intervals for Type B and C

penetrations results in a negligible increase in total plant risk.

Based on the above EOI [Entergy Operation, Inc.] has concluded

that the proposed change will not result in a significant increase

in the probability or consequences of any accident previously

evaluated.

2. The request does not create the possibility of occurrence of

a new or different kind of accident from any accident previously

evaluated.

The proposed change does not involve a change to the plant

design or operation. As a result, the proposed change does not

affect any of the parameters or conditions that could contribute to

initiation of any accidents. This change involves the reduction in

Type B and C test frequency. The methods of performing the tests are

not changed. No new accident modes are created by extending the

testing intervals. No safety-related equipment or safety functions

are altered as a result of this change. Extending

[[Page 56369]]

the test frequency has no influence on , nor does it contribute to, the

possibility of a new or different kind of accident or malfunction

from those previously analyzed.

3. The request does not involve a significant reduction in a

margin to safety.

The proposed change only affects the frequency of Type A, B, and

C testing and does not change the methodology for performance of the

testing. However, the proposed change can increase the probability

that a large increase in leakage could go undetected for an extended

period of time. Operational experience has shown that the leak

tightness of the containment has been maintained significantly below

the allowable leakage limit. In addition, NUREG-1493 has determined

that, under several different accident scenarios, the risk of

radioactivity release from containment is negligible with the

implementation of these proposed changes.

The margin of safety that has the potential of being impacted by

the proposed change involves the offsite [off-site] dose

consequences of postulated accidents which are directly related to

containment leakage rate. The containment isolation system is

designed to limit leakage to La which is defined by the RBS

Technical Specifications to be 0.26 percent by weight of the

containment air per 24 hours at 7.6 psig (Pa). The limitation

on containment leakage rate is designed to ensure that total leakage

volume will not exceed the value assumed in the accident analyses at

the peak accident pressure (Pa) or 7.6 psig. The margin to

safety for the offsite [off-site] dose consequences of postulated

accidents directly related to the containment leakage rate in

maintained by meeting the 1.0 La

No change in the method of testing is being proposed. The Type B

and C tests will continue to be done at full pressure (Pa) or

greater. Other programs are in place to ensure that proper

maintenance and repairs are performed during the service life of the

primary containment and systems and components penetrating the

primary containment.

As a result, EOI had concluded that the proposed change will not

result in a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

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

NRC Project Director: William D. Beckner

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

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

County, New Jersey

Date of amendment request: September 28, 1995

Description of amendment request: The proposed change modifies

Technical Specification 3/4.8.1.2, ``Electrical Power Sources -

Shutdown.'' The surveillance requirement 4.8.1.2 is clarified by a Note

to identify those surveillances which are required to be performed

during Modes 5 and 6.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. Will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

No component modification, system realignment, or change in

operations will occur which could affect the probability of any

accident or transient. The proposed addition of a Note will provide

guidance on which surveillances are required to be performed in

Modes 5 and 6. The Note will preclude rendering operable DGs

inoperable, and/or preclude de-energizing a required ESF bus or

disconnecting a required offsite circuit during the performance of

the surveillance requirement. Proposed changes do not eliminate any

testing requirements, they simply clarify which tests will be

performed in Modes 5 and 6, and which are required to be performed

prior to entry into Mode 4. Therefore, the proposed changes do not

involve a significant increase in the probability or consequences of

an accident previously analyzed.

2. Will not create the possibility of a new or different kind of

accident from any previously evaluated.

No component modification, system realignment, or change in

operating procedure is required to implement the proposed change.

The proposed change reduces the possibility of a single event

impacting the operability of an ESF bus or its DG simultaneously.

Therefore, these changes will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

safety.

The proposed change will not alter any assumptions, initial

conditions, or results of any accident analyses. The Class 1E

equipment assumed available in the accident analyses and their

designed capability to mitigate the consequences of any postulated

accidents will not be changed. The addition of a Note to clarify the

surveillance requirements will not involve a significant reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: John F. Stolz

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

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

County, New Jersey

Date of amendment request: September 28, 1995

Description of amendment request: The proposed changes relocate

``Reactor Coolant System - Chemistry'' Technical Specification 3/4.4.7

(Salem Unit 1) and 3/4.4.8 (Salem Unit 2) and their associated Bases to

the Salem Updated Final Safety Analysis Report (UFSAR) and the

Surveillance Requirements and Limiting Conditions for Operation to

applicable plant procedures controlled by the 10 CFR 50.59 process.

Also, the applicability will be changed from ``At all times'' to

``Modes 1, 2, 3, 4, 5 and 6.''

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. Will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes involve no hardware changes, no changes to

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

existing structures. Specifically, changing the Applicability from

``At all times'' to ``Modes 1, 2, 3, 4, 5 and 6'' by this submittal

will not alter established chemistry for chlorides, fluorides and

dissolved oxygen of the Reactor Coolant System. The relocation of

this Surveillance Requirement/LCOs and Bases to plant procedures and

the UFSAR respectively, will continue to ensure that the chemistry

analysis of the Reactor Coolant System water is monitored and

controlled. Changing the Applicability from ``At all times'' to

``Modes 1,2,3,4,5 and 6'' represent changes that do not affect plant

safety and do not alter existing accident analyses.

2. Will not create the possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes are procedural in nature concerning the

location of the descriptive information and surveillance

requirements for Reactor Coolant System Chemistry. Removing these

specifications from the Technical Specifications and

[[Page 56370]]

placing them in the UFSAR and plant procedures will not alter the

maintenance of the Reactor Coolant System Chemistry or the ability

to monitor its intended functions. Therefore, these changes will not

create a new or unevaluated accident or operating condition.

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

safety.

The proposed changes relocate the Reactor Coolant System

Chemistry Requirements/LCOs from the Technical Specifications to the

UFSAR and plant procedures in accordance with guidance provided by

the NRC Final Policy Statement (58 FR 39132) regarding the

improvement of Technical Specifications. The requirements that will

reside in the UFSAR and plant procedures for the Reactor Coolant

System Chemistry will ensure that the ability to determine chloride,

fluoride and dissolved oxygen concentrations in the Reactor Coolant

System is properly maintained and that the maintenance of the

Reactor Coolant System Chemistry will be commensurate with its

safety significance. Therefore, the proposed changes will not

involve a significant reduction in any margins of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: John F. Stolz

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

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

County, Alabama

Date of amendments request: September 26, 1995

Description of amendments request: The amendments would revise

Technical Specification (TS) Section 4.6.1.3 to incorporate

improvements to containment air lock testing referenced in Chapter 3.6,

``Containment Systems,'' of NUREG-1431, ``Standard Technical

Specifications, WestinghousePlants.''

Basis for proposed no significant hazards consideration

determination: As required by 10CFR 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 change does not involve any change to the

configuration or method of operation of any plant equipment used to

mitigate the consequences of an accident. Containment leakage is an

assumption in the safety analysis of the loss of coolant accident

and the rod ejection accident. Changes to the containment air lock

door seal test acceptance criteria will have no impact on the

radiological consequences of these accidents since the plant safety

analysis is based on the assumption that the containment leaks at

its design leak rate of 0.15 percent per day for the first 24 hours

and 0.075 percent per day thereafter for each of these accidents.

The change to the surveillance requirement meets the intent of the

guidance in NUREG-1431. Primary containment integrity ensures that

the release of radioactive materials from the containment atmosphere

will be restricted to those leakage paths and associated leak rates

assumed in the accident analysis. The limitations on closure and

leak rate for the containment air locks are required to meet these

restrictions on containment integrity. These changes do not increase

the probability that the 10 CFR [Part] 100 limits will be exceeded.

The change to the surveillance requirement does not impose any new

safety analyses limits or alter the plants ability to detect and

mitigate events. Therefore, the proposed change does 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 proposed change involves a revision to the Technical

Specifications to meet the intent of the guidance of NUREG-1431, and

does not necessitate a physical alteration of the plant or change in

parameters governing normal plant operation. The change has not

effect on the plant's compliance with the requirements of Appendix

J. The revision of the acceptance criteria for the air lock door

seal test will improve the FNP [Farley Nuclear Plant] current

testing criteria while maintaining an acceptable level of safety.

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

new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety. The revision of the acceptance criteria of

the air lock door seal test will decrease the overall test burden

without decreasing the margin of safety. The overall leakage rate of

the air lock continues as less than or equal to 0.05La and the

plant safety analysis continues to be based ont he assumption that

the containment leaks at its design leak rate of 0.15 percent per

day for the first 24 hours and 0.075 percent per day thereafter for

each of these accidents. Therefore, the proposed change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201

NRC Project Director: Herbert N. Berkow

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

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

Date of amendment request: July 28, 1995

Description of amendment request: The proposed amendment would

clarify the limiting condition for operation for TS 3.8.1.1 and 3.8.1.2

from ``independent'' circuit to ``qualified'' circuit; explain in the

Bases the requirements for operability of an offsite circuit; delete

the STAGGERED TEST BASIS scheduling requirement to perform emergency

diesel generatorsurveillances; explain 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; remove

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

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

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

circuit to ``qualified'' circuit.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

Toledo Edison has reviewed the proposed changes and determined

that a significant hazards consideration does not exist because

operation ofthe Davis-Besse Nuclear Power Station, Unit No. 1 in

accordance with these changes would:

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

accident previously evaluated because the proposed changes do not

make a change to any accident initiator, initiating condition or

assumption. The proposed changes do not involve a significant change

to the plant design or operation. The proposed changes do not affect

the safety function of the offsite circuits or the emergency diesel

generators (EDGs).

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

accident previously evaluated because the proposed changes do not

invalidate assumptions used in evaluating the radiological

consequences of

[[Page 56371]]

an accident, do not alter the source term or containment isolation and

do not provide a new radiation release path or alter potential

radiological releases.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because the proposed

changes do not introduce a new or different accident initiator or

introduce a new or different equipment failure mode or mechanism.

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

because the proposed changes do not reduce the margin to safety

which exists in the present Technical Specifications [TS] or Updated

Safety Analysis Report. The operability requirements of the TS are

consistent with the initial condition assumptions of the safety

analyses. Further, the proposed changes do not affect the Action

statement requirements for the various levels of degradation in the

offsite [power] circuits or EDGs.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

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

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

NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

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

Date of amendment request: September 29, 1995

Description of amendment request: The proposed amendment would

increase the minimum available borated water volume requirement for the

boric acid addition system, the minimum and maximum boron concentration

requirements for the borated water storage tank, the minimum boron

concentration requirement for the core flood tanks; modify the

surveillance requirements for trisodium phosphate dodecahydrate; and

modify the refueling boron concentration and the associated Action

statement. These proposed changes will affect the following Technical

Specification sections: 3/4.1.2.8, Reactivity Control Systems - Borated

Water Sources - Shutdown; 3/4.1.2.9, Reactivity Control Systems -

Operating; 3/4.5.1, Emergency Core Cooling Systems (ECCS) - Core

Flooding Tanks; 3/4.5.2, Emergency Core Cooling Systems - ECCS

Subsystems - Tavg [plus or minus] 280 deg.F; 3/4.5.4, ECCS - Borated

Water Storage Tank; 3/4.9.1, Refueling Operations - Boron

Concentration; Bases 3/4.1.2, Boration Systems; Bases 3/4.5.2 and 3/

4.5.3, ECCS Subsystems; and Bases 3/4.9.1 Boron Concentration.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

Toledo Edison has reviewed the proposed changes and determined

that a significant hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in

accordance with these changes would:

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

accident previously evaluated because no accident initiators,

conditions, or assumptions are significantly affected by the

proposed changes.

The proposed changes to the Technical Specifications and their

Bases increase the minimum volume of the Boric Acid Addition System

(BAAS), the minimum boron concentration of the Borated Water Storage

Tank (BWST) and Core Flooding Tanks (CFTs), the maximum boron

concentration of the BWST, and the minimum volume of trisodium

phosphate dodecahydrate (TSP) in Containment (CTMT). Administrative

changes to these Technical Specifications have also been proposed.

These changes ensure adequate boration capability is maintained for

normal operations, that adequate Shutdown Margin (SDM) can be

achieved following an accident, and that the assumed post-Loss of

Coolant Accident (LOCA) pH can be achieved. Therefore, as stated

above, these proposed changes do not significantly affect accident

initiators, conditions, or assumptions.

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

accident previously evaluated because the proposed changes do not

change the source term, CTMT isolation, or allowable releases.

In particular, maintaining the appropriate amount of TSP will

ensure the assumed pH will be achieved, the assumption of source

term with respect to iodine retention will be maintained, and the

radiological consequences of a previously evaluated accident will

not be increased.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because no new

accident initiators or assumptions are introduced by the proposed

changes.

These changes ensure that the assumptions used for initial and

final conditions of SDM, pH, and source term are maintained. Also,

the Environmental Qualification (EQ) and seismic requirements have

been verified to be adequate to maintain the adequacy of Structures,

Systems, and Components (SSCs) during assumed accident conditions.

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

because the proposed changes to the minimum volume and boron

concentration for the BAAS, BWST, and CFTs ensure that the margin of

safety for reactor subcriticality is maintained at all times for

future longer fuel cycles, including the upcoming Cycle 11.

The proposed increase in the BWST maximum boron concentration is

set at the conservative limit for post-LOCA boron precipitation

concerns. Therefore, the existing margin of safety with respect to

post-LOCA boron precipitation is maintained.

The proposed increase in the minimum TSP volume requirement

maintains the same margin of safety with respect to post-LOCA pH,

time for dissolution, iodine retention, and chloride stress

corrosion of austenitic stainless steels. The TSP capacity margin of

approximately 40 cubic feet included in the minimum TSP volume

requirement will not result in increasing the pH above the

previously approved pH limit of 11. This reserve capacity adds

margin to ensure adequate minimum pH is achieved.

The proposed removal of the 1800 ppm refueling boron

concentration requirement does not reduce the margin of safety

because the requirement of maintaining keff [less than or equal to]

0.95 is alone sufficient to ensure that the accident analysis

assumptions are satisfied.

The proposed change to the boration rate requirement of the

LCO 3.9.1 Action statement does not reduce the margin of safety

because the proposed boration rate of 12 gpm of 7875 ppm boric acid

solution is equivalent to the present boration rate of

10 gpm of 8750 ppm boric acid solution.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

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

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

NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

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

Date of amendment request: October 2, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Section 5.0, ``Design Features,''

by adding a site location description, remove site area

[[Page 56372]]

maps, remove containment and reactor coolant system design parameters,

remove the description of the meteorological tower location, remove

component cyclic or transient limits, and revise the fuel assembly

description to include the use of ZIRLO clad fuel rods.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

Toledo Edison has reviewed the proposed changes and determined

that a significant hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power Station Unit Number 1, in

accordance with these changes would:

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

accident previously evaluated because no accident initiators,

conditions or assumptions are affected by the proposed changes to

Section 5.0, Design Features, of the Technical Specifications. These

changes are proposed to add a site location description, remove site

area maps, remove containment and reactor coolant system design

parameters, remove the description of the meteorological tower

location, remove component cyclic or transient limits, and revise

the fuel assembly description to include the use of ZIRLO clad fuel

rods.

Under the proposed changes, Technical Specifications (TS)

Section 5.0 would continue to satisfy the applicable requirements of

Section 182.a of the Atomic energy Act of 1954, and 10 CFR

50.36(c)(4). Further, the proposed changes are consistent with

NUREG-1430, ``Standard Technical Specifications for Babcock and

Wilcox Plants,'' Revision 1. The information proposed for removal

from existing TS 5.0 is presently included in the Updated Safety

Analysis Report (USAR) or is being proposed to be added to the USAR,

hence sufficient controls exist under 10 CFR 50.59 to ensure that

future changes to these items are acceptable.

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

accident previously evaluated because no accident conditions or

assumptions are affected by the proposed changes. As described

above, these changes are consistent with the ``Standard Technical

specifications for Babcock and Wilcox Plants'' (NUREG-1430) and are

administrative changes. The proposed changes do not alter the source

term, containment isolation, or allowable releases. The proposed

changes, therefore, will not increase the radiological consequences

of a previously evaluated accident.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because no new

accident initiators or assumptions are introduced by the proposed

changes, which involve only administrative controls. As described

above, these changes are consistent with the ``Standard Technical

Specifications for Babcock and Wilcox Plants'' (NUREG-1430) and are

administrative changes. The proposed changes do not alter any

accident scenarios.

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

because the proposed changes are administrative and do not reduce or

adversely affect the capabilities of any plant structure, systems or

components.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

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

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

NRC Project Director: Gail H. Marcus

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: September 6, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 5.3.1 to reflect a change in the

maximum initial enrichment for reload fuel. The amendment would also

change the maximum reference Kinfinity for storage in Region 1 of

the spent fuel pool and TS Figure 3.9-1 to reflect a change in the

maximum initial enrichment for storage in Region 2.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

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

the probability or consequences of an accident previously evaluated.

An increase to a maximum initial enrichment of 5.0 w/o U-235

does not involve an increase in the probability or consequence of an

accident or other adverse condition over previous evaluations.

Because of the conservative techniques and assumptions used to

evaluate the maximum possible neutron multiplication factor, there

is reasonable assurance that criticality safety is maintained when

storing fuel assemblies of up to and including 5.0 w/o U-235 in the

spent fuel storage racks under both normal and postulated accident

conditions. For example, the calculations for non-accident

conditions ignore the 2000 ppm soluble boron in the spent fuel pool

calculations, thus resulting in conservative values of the

multiplication factor. Storing fuel in the Region 1 configuration

which meets the IFBA [integral fuel burnable absorber] versus

enrichment curve (Figure 3 of Attachment 6) results in a maximum

multiplication factor of 0.9481, including all biases and

uncertainties.

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

or different kind of accident from any previously evaluated.

An increase to a maximum initial enrichment level of 5.0 w/o U-

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

accident or condition over previous evaluations. An increase to the

enrichment level of 5.0 w/o U-235 involved performing extensive

evaluations to develop the IFBA versus enrichment curve for V-5

fuel. Use of dual code packages ensures that the spent fuel pool

Region 1 criticality limits are not exceeded.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

An increase in the maximum initial enrichment level to 5.0 w/o

U-235 does not involve a reduction in the margin of safety. As

discussed above, in all cases the multiplication factors for worst

case assumptions fall considerably below the criticality limits and

do not represent any reductions in margin. An increase to the

initial enrichment level of 5.0 w/o U-235 does not adversely impact

operation of the various plant systems, i.e. HVAC [heating,

ventilation, and air conditioning], spent fuel pool cooling, or

radiological control systems.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

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

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

NRC Project Director: William H. Bateman

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: October 6, 1995

Description of amendment request: The proposed amendment would

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

4.2.b, ``Steam Generator Tubes,'' its associated bases, and Figure TS

4.2-1 by redefining the pressure boundary for Westinghouse mechanical

hybrid expansion joint (HEJ) steam generator (SG) tube sleeves.

Basis for proposed no significant hazards consideration

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

[[Page 56373]]

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

The proposed change was reviewed in accordance with the

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

1. Operation of the KNPP in accordance with the proposed license

amendment does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

Mechanical testing has shown that the inherent structural

strength of the HEJ joint provides sufficient integrity such that

the tube rupture capability recommendations of RG [Regulatory Guide]

1.121 are met, even for instances of 100 percent throughwall,

360 deg. circumferentially oriented degradation in the HEJ HRLT

[hardroll lower transition] region. Structural integrity

recommendations consistent with RG 1.121 are supplied

for all tube degradation 1.1 inch or greater below the bottom of

the HEJ HRUT [hardroll upper transition]. Based on test data, a

bounding SLB [steam line break] leak rate of 0.033 gpm for

indications between 1.1 and 1.3 inch below the bottom of the HRUT is

applied. As the leakage data base is expanded and statistical basis

established, this SLB leakage allowance may be reduced. For

indications existing greater than 1.3 inch below the bottom of the

HRUT, SLB event leakage can be neglected.

Additional prevention from tube rupture is inherently provided

by the HEJ geometry. For RCS [reactor coolant system] release rates

to exceed the normal makeup capacity of the plant, the tube must be

postulated to experience a complete circumferential separation at

the lower transition, and become axially displaced by 3 to 3.25

inches, resulting in complete geometric disassociation between the

tube and sleeve resulting in sufficient flow area to support leakage

in excess of makeup capacity. During the 1989 plug top release event

at North Anna Unit 1, primary to secondary release rates were

calculated to be less than 80 gpm, for a flow area approximately

four times larger than the flow area created by a tube which was

axially displaced by about 1.25 to 1.5 inch. Analysis of the steam

generator indicates that at a 95 percent cumulative probability, the

tube would experience an axial displacement of less than the 1.1

inch boundary. At this level of axial displacement, a ring of metal

to metal contact would remain between the tube and sleeve, and

leakage would be far less than makeup. Projected leakage at this

point is expected to be less than 2.5 gpm. Therefore, implementation

of the proposed repair boundary will not result in tube rupture,

even for a tube postulated to not behave as predicted by the

available test and pulled tube data.

The proposed technical specification change to support the

implementation of the HEJ sleeve tube pressure boundary for parent

tube degradation in the HEJ HRLT region does not adversely impact

any other previously evaluated design basis accident or the results

of accident analyses for the current technical specification minimum

reactor coolant system flow rate. Plugging limit criteria are

established using the guidance of RG 1.121. Furthermore, per RG 1.83

recommendations, the sleeved tube assembly can be monitored through

periodic inspections with present eddy current techniques.

2. The proposed license amendment does not create the

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

previously evaluated.

Implementation of the revised pressure boundary will not

introduce significant or adverse changes to the plant design basis.

Mechanical testing of degraded sleeve joints supports the

conclusions of the calculations that the sleeve retains structural

(tube burst) capability consistent with RG 1.121. As with initial

installation of sleeves, implementation of the relocated pressure

boundary cannot interact with other portions of the RCS. Any

hypothetical accident as a result of potential tube degradation in

the HEJ HRLT region of the tube is bounded by the existing tube

rupture accident analysis. Neither the sleeve design nor

implementation of the tube repair boundary defined on Figure TS 4.2-

1 affects any other component or location of the tube outside of the

immediate area repaired.

3. The proposed license amendment does not involve a significant

reduction in a margin of safety.

The safety factors used in the establishment of the HEJ sleeved

tube pressure boundary are consistent with the safety factors in the

ASME [American Society of Mechanical Engineers] Boiler and Pressure

Vessel Code used in steam generator design. Based on the sleeved

tube geometry, it is unrealistic to consider that application of the

revised pressure boundary could result in single tube leak rates

exceeding the normal makeup capacity during normal operating

conditions. The pressure boundary established ... has been developed

using the methodology of RG 1.121. The performance characteristics

of postulated degraded parent tubes of HEJ tube/sleeve joints have

been verified by testing to retain structural integrity and preclude

significant leakage during normal and postulated accident

conditions. Testing indicates that postulated circumferentially

separated tubes which the repair boundary addresses would not

experience axial displacement during either normal operation or SLB

conditions. The existing offsite dose evaluation performed for KNPP

in support of the voltage based plugging criteria for axial ODSCC

[outside diameter stress corrosion cracking] at TSP [tube support

plate] intersections established a faulted loop primary to secondary

leak rate of 34.0 gpm using technical specification dose equivalent

Iodine-131 activity levels. Following implementation of the

criteria, postulated leakage from all sources must not exceed 34.0

gpm in the faulted loop. Maintenance of this limit will ensure that

offsite doses would not exceed the currently accepted limit of a

small fraction of the 10 CFR 100 guidelines. The repair boundary

uses a conservatively established ``per indication'' leak rate for

estimation of SLB leakage. This leak rate is applied to all

indications left in service as a result of the tube repair boundary,

including non-throughwall indications and a limited number of

indications of circumferential throughwall extent.

For a postulated indication whose performance is not

characteristic of the test and pulled tube data, and which would

experience axial displacement at the 95 percent cumulative

probability value following a postulated SLB event with no operator

intervention, leakage would not be expected to result in an

uncontrolled release of reactor coolant in excess of normal makeup

capacity.

For the three removed tube sleeve samples and nearly 1,000 PTIs

[parent tube indications] detected in the field, there were no

instances of degradation of elevations (multiple expansion

transitions) on either side of the hardroll expansion in the same

tube. This includes no instances on non-detected degradation in the

upper hydraulic and hardroll upper expansion transitions for the

removed tubes. One tube was identified in the most recent KNPP

inspection with two separate circumferential crack elevations within

the HRLT. Rapidly occurring degradation would not be expected at the

upper transitions, based partly on the field inspection results. The

available inspection results include two inspection programs (1994

and 1995) at Kewaunee and one at Point Beach Unit 2 (1994). Through

these three inspection programs, approximately 11,000 HEJ sleeved

tubes have been inspected using advanced ET [eddy current testing]

techniques.

The portions of the installed sleeve assembly which represent

the reactor coolant pressure boundary can be monitored for the

initiation and progression of sleeve/tube wall degradation, thus

satisfying the requirements of Regulatory Guide 1.83.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

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

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

NRC Project Director: Gail H. Marcus

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments

[[Page 56374]]

issued or proposed to be issued involving no significant hazards

consideration.

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

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

of the original notice.

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

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

Date of amendment request: October 6, 1995

Description of amendment request: Revise the Technical

Specifications to change the definition of the F* distance.

Date of publication of individual notice in Federal Register:

October 16, 1995 (60 FR 53648)

Expiration date of individual notice notice: November 15, 1995

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085

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

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

Manitowoc County, Wisconsin

Date of application for amendment: September 13, 1995, as

supplemented by letter dated October 19, 1995

Brief description of amendment request: The proposed amendments

would revise Technical Specification (TS) Section 15.1,

``Definitions,'' the basis for TS Section 15.3.1.G, ``Operational

Limitations,'' and TS Figure 15.2.1-2, ``Reactor Core Safety Limits,

Point Beach Unit 2.'' The proposed changes would reduce the reactor

coolant system raw measured total flow rate limit and reflect new

reactor core safety limits for Unit 2.Date of individual notice in

Federal Register: October 24, 1995 (60 FR 54527)

Expiration date of individual notice notice: November 8, 1995

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

Sixteenth Street, Two Rivers, Wisconsin

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

County, Maryland

Date of application for amendments: June 6, 1995

Brief description of amendments: The amendments extend the nominal

surveillance interval requirements of selected safety systems

instruments form 18 months to a refueling interval of 24 months.

Date of issuance: October 19, 1995

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

30 days for Unit 2 and prior to restart of the spring 1996 refueling

outage for Unit 1.

Amendment Nos.: 208 and 186

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 5, 1995 (60 FR

35061) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated October 19, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Calvert County Library,

Prince Frederick, Maryland 20678

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

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

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

Will County, Illinois Docket Nos. 50-10, 50-237 and 50-249, Dresden

Nuclear Power Station, Units 1, 2 and 3, Grundy County, Illinois

Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and

2, LaSalle County, Illinois Docket Nos. 50-254 and 50-265, Quad

Cities Nuclear Power Station, Units 1 and 2, Rock Island County,

Illinois Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station,

Units 1 and 2, Lake County, Illinois

Date of application for amendments: April 24, 1995, as supplemented

August 1 and September 14, 1995.

Brief description of amendments: The amendments would relocate the

requirements for the ``Review, Investigative and Audit Functions'' and

frequencies of the quality assurance (QA) program from the

administrative controls section of the TS to the appropriate sections

of the licensee's Quality Assurance Topical Report (QATR), CE-1-A,

Revision 65. In addition, the proposed TS changes include title changes

to reflect the reorganization of the licensee's Nuclear Operations

Division and miscellaneous administrative and editorial changes.

Date of issuance: October 20, 1995

Effective date: October 20, 1995

Amendment Nos.: 75, 75, 67, 67, 38, 141, 135, 107, 93, 163, 159,

171, and 158

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

DPR-2, DPR-19, DPR-25, NPF-11 NPF-18, DPR-29, DPR-30, DPR-39 and DPR-

48: The amendments revised the Technical Specifications.

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

45175) and September 20, 1995 (60 FR 48726). The August 1 and September

14, 1995, letters provided clarifying information that did not change

the initial proposed no significant hazards consideration determination

or expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 20, 1995.No significant hazards

consideration comments received: No

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

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

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481; for Dresden, Morris Area Public Library

District, 604

[[Page 56375]]

Liberty Street, Morris, Illinois 60450; for LaSalle, Jacobs Memorial

Library, Illinois Valley Community College, Oglesby, Illinois 61348;

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

Illinois 61021; and for Zion, Waukegan Public Library, 128 N. County

Street, Waukegan, Illinois 60085

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

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

Date of application for amendments: June 30, 1995

Brief description of amendments: The amendments modify the

surveillance requirements for the emergency diesel generators.

Date of issuance: October 16, 1995

Effective date: October 16, 1995

Amendment Nos.: 170 and 157

Facility Operating License Nos. DPR-39 and DPR-48: The amendments

revised the Technical Specifications.

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

FR 47615) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 16, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085

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

Unit No. 2, Pope County, Arkansas

Date of amendment request: April 4, 1995

Brief description of amendment: The amendment deletes requirements

associated with part length control element assemblies.

Date of issuance: October 12, 1995

Effective date: October 12, 1995

Amendment No.: 169

Facility Operating License No. NPF-6. Amendment revised the

Technical Specifications.

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

37090) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 12, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Unit No. 2, Pope County, Arkansas

Date of application for amendment: March 17, 1995

Brief description of amendment: The amendment deletes requirements

associated with surveillance to verify position stops for High Pressure

Safety Injection Emergency Core Cooling System throttle valves.

Date of issuance: October 18, 1995

Effective date: October 18, 1995

Amendment No.: 170

Facility Operating License No. NPF-6. Amendment revised the

Technical Specifications.

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

37089) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 18, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Unit No. 2, Pope County, Arkansas

Date of amendment request: April 4, 1995, as supplemented by letter

dated October 12, 1995

Brief description of amendment: The amendment revises the

containment cooling response time to reduce the likelihood of a water

hammer event in service water piping.

Date of issuance: October 26, 1995

Effective date: October 26, 1995

Amendment No.: 171

Facility Operating License No. NPF-6. Amendment revised the

Technical Specifications.

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

37090) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 26, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

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

Date of application for amendments: February 28, 1994

Brief description of amendments: The amendments delete the minimum

frequency criteria prescribed for quality assurance audits from

Administrative Controls sections 6.5.2.8 and 6.8.4 of the Technical

Specifications (TS). Audit periodicity will thereby be controlled by

the program described in the Florida Power and Light Company (FPL)

Topical Quality Assurance Report.

Date of issuance: October 25, 1995

Effective date: October 25, 1995

Amendment Nos.: 140 and 80

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the Technical Specifications.

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

17599) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 25, 1995. No significant

hazards consideration comments received: No

Local Public Document Room location: Indian River Junior College

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

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

Brief description of amendments: These amendments consist of

administrative corrections and clarifications.

Date of issuance: October 17, 1995

Effective date: October 17, 1995

Amendment Nos. 177 and 171Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

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

FR 47619) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 17, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

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

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

Date of application for amendments: July 26, 1995

Brief description of amendments: These amendments consist of

administrative corrections and clarifications.

Date of issuance: October 17, 1995

Effective date: October 17, 1995

Amendment Nos.: 178 and 172Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

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

FR 47619) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 17, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

[[Page 56376]]

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

Linn County, Iowa

Date of application for amendment: February 13, 1995, as

supplemented April 21, 1995, and August 7, 1995.

Brief description of amendment: The proposed amendment deletes the

audit requirements from the Duane Arnold Energy Center Technical

Specifications (TS) and adds them to the Quality Assurance Program.

Date of issuance: October 17, 1995

Effective date: October 17, 1995

Amendment No.: 213

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

Technical Specifications.

Date of initial notice in Federal Register: March 29, 1995 (60 FR

16190) The additional information contained in the supplemental letters

dated April 21, 1995, and August 7, 1995, was clarifying in nature and

did not change the NRC staff's initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 17, 1995.No significant hazards

consideration comments received: No

Local Public Document Room location: Cedar Rapids Public Library,

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

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 application for amendments: March 31, 1995

Brief description of amendments: The amendments revise Technical

Specification (TS) surveillance requirements for safety-related pump

testing to eliminate recirculation alignments. In addition, specific

test parameters, discharge pressures, and flows associated with these

pumps are removed from the TS and will be controlled by the Inservice

Testing Program.

Date of issuance: October 17, 1995

Effective date: October 17, 1995, with full implementation within

45 days

Amendment Nos.: 203 and 188

Facility Operating License Nos. DPR-58 and DPR-74. Amendments

revised the Technical Specifications.

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

32368) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 17, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

Date of application for amendment: February 1, 1995

Brief description of amendment: The amendment revises Technical

Specification 3.6.13 and associated Bases to permit the controls and

instruments from both Remote Shutdown Panels to be considered when

assuring that one complete set of controls and instruments is operable.

The changes also allow 30 days to restore an inoperable function to

operable status, remove MODE 3 (hot shutdown) from the existing

requirement for operability, and revise the LIMITING CONDITION FOR

OPERATION ACTION to require achieving hot shutdown in 12 hours instead

of cold shutdown in 36 hours. An additional change permits the operator

30 days to establish an alternate method of monitoring a parameter (and

90 days to restore the function) when the function is inoperable.

Date of issuance: October 16, 1995

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

30 days.

Amendment No.: 155

Facility Operating License No. DPR-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: March 1, 1995 (60 FR

11135) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 16, 1995. No significant

hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

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

York 13126

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

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

Connecticut

Date of application for amendment: March 29, 1995

Brief description of amendment: The amendment modifies the current

Technical Specifications that have cycle-specific parameter limits in

the Core Operating Limits Report to include an additional cycle-

specific parameter and its supporting methodologies.

Date of issuance: October 18, 1995

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

60 days.

Amendment No.: 120

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

Technical Specifications.

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

24912) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 18, 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

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit

No. 3, York County, Pennsylvania

Date of application for amendment: September 1, 1995

Brief description of amendment: The amendment deleted License

Condition 2.C.(5) which restricts power levels to no less than seventy

percent in the coastdown condition.

Date of issuance: October 17, 1995

Effective date: As of date of issuance

Amendment No.: 215

Facility Operating License No. (DPR-56): This amendment revised the

Facility Operating License. Public comments requested as to proposed no

significant hazards consideration: Yes. (60 FR 48530). That notice

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

no significant hazards consideration determination. No comments have

been received. The notice also provided for an opportunity to request a

hearing by October 18, 1995, but indicated that if the Commission makes

a final no significant hazards consideration determination any such

hearing would take place after issuance of the amendment.

The Commission's related evaluation of the amendment, finding of

exigent circumstances, and final no significant hazards consideration

determination are contained in a Safety Evaluation dated October 17,

1995.

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. Vice

President and General Counsel, PECO Energy Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

Local Public Document Room location: Government Publications

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

[[Page 56377]]

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

Pennsylvania 17105

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

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

Pennsylvania

Date of application for amendments: May 19, 1995

Brief description of amendments: The amendments revise the

Technical Specifications Table 3.3.3-3, ``Emergency Core Cooling System

Response Times'' to reflect the value of 60 seconds for the High

Pressure Coolant Injection system response time instead of 30 seconds

as previously specified.

Date of issuance: October 16, 1995

Effective date: For both units, as of the date of issuance and to

be implemented within 30 days.

Amendment Nos.: 102 and 66

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 5, 1995 (60 FR

35084) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 16, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464

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

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

Date of application for amendment: July 21, 1995

Brief description of amendment: The amendment revises TS Section

6.0 (Administrative Controls) to replace the title-specific list of

members on the Plant Operating Review Committee with a more general

statement of membership requirements, and expands the scope of

disciplines represented on the committee to include Nuclear Licensing

and Quality Assurance. The amendment also changes the following

management position titles: ``First Executive Vice President and Chief

Nuclear Officer'' to ``Chief Nuclear Officer'', ``Resident Manager'' to

``Site Executive Officer'', ``Shift Supervisor'' to ``Shift Manager'',

and ``Assistant Shift Supervisor'' to ``Control Room Supervisor.''

These changes in title do not affect the reporting relationships,

authority, or responsibilities of these positions. Finally, the

amendment also makes editorial corrections to the TSs.

Date of issuance: October 13, 1995

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

30 days.

Amendment No.: 228

Facility Operating License No. DPR-59: 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 October 13, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

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

York 13126

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

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

Date of application for amendment: April 12, 1995.

Brief description of amendment: The amendment extends the

surveillance test intervals for the nuclear steam supply system to

support 24-month operating cycles. Surveillance test interval

extensions that are justified will be denoted as being performed

``every 24 months'' or ``at least once per 24 months'' consistent with

the guidance provided in Reference 1. Other surveillances currently

performed ``once each operating cycle,'' ``at least once during each

operating cycle,'' ``each refueling,'' or similar notation, that are

not being extended at this time will be denoted as being performed ``at

least once per 18 months.'' The NRC staff has determined that the

proposed TS changes follow the guidance of Generic Letter 91-04, and

are therefore acceptable.

Date of issuance: October 13, 1995

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

30 days.

Amendment No.: 229

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

Technical Specifications.

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

24916) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 13, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

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

York 13126

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

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: April 18, 1995

Brief description of amendment: This amendment changes Technical

Specification Table 4.3.7.1-1, ``Radiation Monitoring Instrumentation

Surveillance Requirements,'' to increase the channel functional test

interval from monthly to quarterly for each instrument.

Date of issuance: October 16, 1995

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

within 60 days.

Amendment No.: 83

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

Technical Specifications.

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

42607) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 16, 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. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: May 4, 1995

Brief description of amendment: This amendment changes Technical

Specification (TS) 3/4.6.1.8, ``Drywell and Suppression Chamber Purge

System,'' increasing the annual operational limit for the drywell and

suppression chamber purge system from 120 to 500 hours.

Date of issuance: October 16, 1995

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

within 60 days.

Amendment No.: 84

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

Technical Specifications.

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

42607) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 16, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

[[Page 56378]]

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

Creek Generating Station, Salem County, New Jersey

Date of applications for amendment: November 30, 1994 and March 30,

1995, as supplemented by letter dated September 5, 1995.

Brief description of amendment: The change to TS Table 3.3.1-2,

``Reactor Protection System Response Times,'' TS Table 3.3.2-3,

``Isolation System Instrumentation Response Time,'' TS Table 3.3.3-3,

``Emergency Core Cooling System Response Times,'' and associated Bases,

eliminates the requirement to perform response time testing for certain

classes of equipment and transfers the requirements of the above-

referenced TS Tables to the Updated Final Safety Analysis Report.

Date of issuance: October 24,1995

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment No.: 85

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

Technical Specifications.

Date of initial notice in Federal Register: March 29, 1995 (60 FR

16198 and August 16, 1995 (60 FR 42606) The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

October 24, 1995. No significant hazards consideration comments

received: No

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment: May 20, 1994, as supplemented on

March 29, 1995

Brief description of amendment: The amendment revises Technical

Specifications to implement the NRC's Final Policy Statement on

Technical Specification Improvements for Nuclear Power Reactor by

relocating specifications that do not meet policy statement criteria to

the Final Safety Analysis Report.

Date of issuance: October 20, 1995

Effective date: Immediately, to be implemented within 120 days.

Amendment No.: 103

Facility Operating License No. NPF-30. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45036). The March 29, 1995, letter provided supplemental information

that did not change the initial proposed no significant hazards

consideration determination or expand the scope of the original Federal

Register notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 20, 1995.No significant hazards

consideration comments received: No

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

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

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

Manitowoc County, Wisconsin

Date of application for amendments: January 26, 1994, as

supplemented by letters dated December 1, 1994, and June 23, 1995

Brief description of amendments: These amendments revise Technical

Specification (TS) Section 15.3.0, ``General Considerations.'' This

section specifies the actions to be taken for conditions not directly

addressed in the action statements fo the TSs. In addition, changes to

the applicable bases (including the bases for TS 15.3.3) and editorial

changes are also included.

Date of issuance: October 12, 1995

Effective date: October 12, 1995

Amendment Nos.: 163 and 167

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

revised the Technical Specifications.

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

12373) The December 1, 1994 and June 23, 1995, submittals provided

supplemental information that did not change the initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 12, 1995.No significant hazards

consideration comments received: No

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

Sixteenth Street, Two Rivers, Wisconsin 54241

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

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

Manitowoc County, Wisconsin

Date of application for amendments: April 17, 1995

Brief description of amendments: These amendments change TS

Sections 15.6.2, ``Organization,'' and 15.6.3, ``Facility Staff

Qualifications.'' The requirement for the Operations Manager to hold an

NRC Senior Reactor Operator's (SRO) license has been changed to provide

additional staffing flexibility.

Date of issuance: October 12, 1995

Effective date: October 12, 1995

Amendment Nos.: 164 and 168

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

revised the Technical Specifications.

Date of initial notice in Federal Register: May 23, 1995 (60 FR

27346). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 12, 1995.No significant

hazards consideration comments received: No

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

Sixteenth Street, Two Rivers, Wisconsin 54241

Dated at Rockville, Maryland, this 1st day of November 1995.

For the Nuclear Regulatory Commission

Jack W. Roe,

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

Reactor Regulation

[Doc. 95-27543 Filed 11-7-95; 8:45 am]

BILLING CODE 7590-01-F