Applications And Amendments To Facility Operating Licenses Involving No Significant Hazards Considerations

[Federal Register Volume 62, Number 98 (Wednesday, May 21, 1997)]

[Notices]

[Pages 27792-27810]

From the Federal Register Online via the Government Publishing Office [www.gpo.gov]

[FR Doc No: X97-10521]

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications And Amendments To Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from April 28, 1997 through May 9, 1997. The last

biweekly notice was published on May 7, 1997 (62 FR 24984).

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

[[Page 27793]]

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 Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, 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 June 20, 1997, 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 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 20555-0001, 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 20555-0001,

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).

[[Page 27794]]

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

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

Date of amendments request: December 27, 1996

Description of amendments request: The proposed amendments would

revise Technical Specification (TS) 3.6.1.3.b (peak containment

internal pressure for the design basis loss of coolant accident (LOCA))

from 49.5 psig to 52 psig and the associated Bases Sections. The

proposed amendments reflect values based on a revised LOCA analysis.

The LOCA analysis was revised to reflect the maximum primary

containment internal pressure specified in other TS. This maximum

primary containment internal pressure was not used in the original LOCA

analysis.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff's analysis is presented below.

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed amendment increases the peak calculated

containment internal pressure for the design basis LOCA from 49.5

psig to 52 psig. The maximum pressure occurs following an accident.

Since the pressure is a consequence of an accident, this change has

no effect on the probability of accident initiation, and therefore,

the probability of an accident previously evaluated has not been

significantly increased.

The consequences of an accident previously evaluated in the

Updated Final Safety Analysis Report (UFSAR) will not be

significantly increased. UFSAR Section 15.6.5.6, ``Analyses of

Effects and Consequences - Large Break LOCA,'' states that ``It is

assumed that the containment leaks at the maximum rates allowed by

the Technical Specifications, i.e., 0.1 vol. %/d for the first 24

hours and half of that rate thereafter.'' The dose calculation

assumes that under accident conditions, the release of radionuclides

to the containment is instantaneously homogenized within the

containment free air volume. This results in a constant

radioactivity per volume (curies/cc) regardless of containment

internal pressure. Since radioactivity is assumed to be homogenized

in the containment free air volume, the volume percent leaked per

day is equivalent to the fraction of radioactivity which leaks from

the containment per day. Therefore, the increase in the peak

calculated containment internal pressure for the design basis LOCA

from 49.5 psig to 52 psig does not effect dose consequences

associated with the design basis LOCA. The proposed change to the

peak calculated containment internal pressure for the design basis

LOCA does not impact the radiological consequences of a LOCA as

analyzed in Chapters 6 and 15 of the UFSAR.

The proposed amendments do not, therefore 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 possibility of a new or different kind of accident has not been

created. The increase in the peak calculated containment internal

pressure for the design basis LOCA does not affect the design or

operation of existing plant equipment, nor involve new plant equipment.

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 containment design pressure is 60 psig. The acceptance

criteria in NRC Standard Review Plan, Section 6.2.1.1.A, ``PWR Dry

Containments, including Subatmospheric Containments,'' requires in

Item 11.1 that ``the containment design pressure should provide at

least a 10% margin above the accepted peak calculated containment

pressure following a loss of coolant accident.'' For PVNGS to

maintain the required margin, this requires that the peak calculated

containment internal pressure for the design basis LOCA would be no

higher than 54 psig. Since the revised peak calculated containment

internal pressure for the design basis LOCA remains below the 54

psig limit, the proposed change does not involve a significant

reduction in the margin of safety.

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

CFR50.92(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

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

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

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

Date of amendment request: April 9, 1997

Description of amendment request: The proposed change will extend

the existing Technical Specifications surveillance intervals from 7

days to 14 days for the Channel Functional Tests for the refueling

equipment interlocks and for the one-rod-out interlock. The change will

permit, under most normal circumstances, a complete offloading,

shuffling, or onloading of fuel, without the need to halt refueling

activities solely for the performance of these surveillance tests.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change extends the Technical Specification

Surveillance Requirement (SR) Frequency for the Channel Functional

Tests (CFTs) for the refueling equipment interlocks and the one-rod-

out interlock. The refueling equipment interlocks and the one-rod-

out interlock are explicitly assumed in the analysis of the control

rod removal error during refueling. Criticality, and therefore,

subsequent prompt reactivity excursions are prevented during the

insertion of fuel, provided all control rods are fully inserted

during the fuel insertion. The refueling equipment interlocks

accomplish this by preventing loading fuel into the core with any

control rod withdrawn, or by preventing withdrawal of a control rod

from the core during fuel loading. The one-rod-out interlock and

adequate shutdown margin prevent criticality by preventing

withdrawal of more than one control rod. With one control rod

withdrawn, the core will remain subcritical, thereby preventing any

prompt critical excursion. The proposed change does not change the

function of any of these interlocks, only the frequency at which the

interlocks undergo channel functional testing. A review of past test

performances has demonstrated that extending the Frequency from 7

days to 14 days will not result in any increase in test failures.

Therefore, the proposed change will not change the ability of these

interlocks to perform when required. Based on this, there can be no

significant increase in the radiological consequences of any

previously evaluated accident since all interlocks will continue to

perform as presently analyzed. Therefore, the proposed change does

not involve a significant increase in the

[[Page 27795]]

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any previously evaluated.

The proposed change extends the SR Frequency for performing CFTs

for refueling equipment and one-rod-out interlocks. This change does

not result in a modification to the plant or to the manner in which

the plant is operated. The testing will still demonstrated the

operability of the interlocks. Thus, the interlocks will still

function in the same manner. Therefore, the proposed change does not

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

any previously evaluated.

3. The proposed change will not involve a significant reduction

in the margin of safety.

The proposed change extends the SR Frequency for performing CFTs

on the refueling equipment and one-rod-out interlocks from 7 days to

14 days. Reviews of past test results indicate that extending the

test interval to 14 days will not result in an increase in the

number of CFT failures for these interlocks. This implies that

extending the SR Frequency to 14 days will not result in an increase

in the amount of time the instrument channels will be inoperable

when required to be operable. Since the proposed change does not

result in any reduction in the amount of time the instrument

channels will be operable, the proposed change does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

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

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

NRC Project Director: Gail H. Marcus

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

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

Date of amendment request: April 21, 1997

Description of amendment request: The proposed amendment would

revise the Technical Specifications that would (1) reduce the volume of

borated water in the core flood tank (CFT) from 1040 cubic feet to 940

cubic feet, (2) reduce the surveillance acceptance criteria for the

emergency core cooling system (ECCS) high pressure injection (HPI)

flowrate from 500 gallons per minute (GPM) to 431 GPM, and (3) revise a

limiting condition for operation (LCO) which currently allows either

local or remote manual operability of decay heat valves to delete the

local manual valve operability option.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

1. State the basis for the determination that the proposed

activity will not represent a significant increase in the

probability of occurrence or consequences of an accident.

This TSCR [Technical Specification change request] revises the

LCO for RB [reactor building] sump isolation valves, the LCO for the

core flood tank level, and the surveillance requirement for HPI

injection flow rate. The Core Flood and HPI systems are not actuated

until an event occurs. The CFT level used in the new accident

analysis is that level required to be maintained in the CFT

throughout operation (i.e., pre-accident). The new CFT level does

not prevent safe accident mitigation.

Likewise, the reduced HPI flow cannot cause an event to occur,

and while such flow results in less injection to the RCS [reactor

coolant system] when actuated, this is acceptable as demonstrated in

the LOCA [loss-of-coolant accident] analyses. Changes to the LCO for

the RB sump isolation valves support the safety analysis

assumptions. The action statements related to both the level

requirement and flow rates remain unchanged by this request. The

function, operation and surveillance intervals for the isolation

valves (DH-V-6A/B), the CFT level and HPI injection system are not

changed by this request. Therefore, this activity does not increase

the probability of occurrence of an accident, previously evaluated

in the SAR [safety analysis report].

Reducing the CFT nominal volume and reducing the HPI flow

acceptance criteria in the Technical Specifications will not

increase the radiological consequences of any LOCA evaluated in the

SAR. The results of analyses using the reduced CFT inventory and

reduced HPI flow demonstrate that the consequences are within the

limits of 10 CFR 50.46. No fuel failure in addition to that assumed

in the evaluation of the dose consequences would occur. Therefore,

the radiological consequences would not increase.

The editorial changes described above have no impact upon the

probability of occurrence or consequences of an accident.

2. State the basis for the determination that the activity does

not create the possibility of an accident of a new or different type

than any previously analyzed in the SAR.

This TSCR revises the LCO for RB sump isolation valves, the LCO

for the core flood tank level, and the surveillance requirement for

HPI injection flow rate. This change will not adversely affect the

capability of the emergency core cooling systems in the event of a

LOCA. The function, operation and surveillance intervals for both

the borated water level in the core flood tank, and ECCS systems are

not changed by this request and no physical changes or modifications

are being made to Core Flood and HPI system boundaries. Therefore,

because there are no configuration changes this activity does not

create the possibility of an accident or malfunction of a different

type than previously analyzed in the SAR.

In addition, the editorial changes described above do not create

the possibility of an accident of a new or different type than any

previously analyzed in the SAR.

3. State the basis for the determination that the margin of

safety is not significantly reduced.

This TSCR revises the LCO for RB sump isolation valves, the LCO

for the core flood tank level, and the surveillance requirement for

HPI injection flow rate. No system configuration changes (hardware

modifications) will be made to implement the change request, upon

approval of the license amendment. The action requirements for these

technical specifications have not changed. Actions to be taken if

operability requirements are not met include plant shutdown under

certain conditions.

Furthermore, impact upon the margin to safety is limited because

the results of the LOCA analyses demonstrate that the 10 CFR 50.46

acceptance criteria are met, specifically: the PCT [peak clad

temperature] limit and the core-wide oxidation limit of 1 percent of

the fuel cladding, as identified in the Technical Specification

bases. Hence the margin of safety as defined in the bases of any

technical specification is not significantly reduced or impacted by

the implementation of this change request, or the editorial changes

described above.

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: Law/Government Publications

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

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

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

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

NRC Project Director: Patrick D. Milano, Acting

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: April 22, 1997

Description of amendment request: The proposed amendments would

revise Technical Specifications 5.3.1, Fuel Assemblies, and 6.9.1.6,

Core Operating Limits Report, to allow use of

[[Page 27796]]

an alternate zirconium-based fuel cladding, ZIRLO, and limited

substitution of fuel rods by ZIRLO filler rods.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The methodologies used in the accident analyses remain

unchanged. With the exception of a reduction in the heat flux hot

channel factor (FQ), the operating limits will not be

changed. The proposed changes will not result in any equipment

exceeding its design limits under normal or accident conditions. The

calculated doses presented in the UFSAR will remain bounding. Other

than the changes to the fuel assemblies, there are no physical

changes to the plant associated with this Technical Specification

change. A reload safety analysis will continue to be performed for

each cycle to demonstrate compliance with fuel safety design bases.

VANTAGE+ fuel assemblies with ZIRLO clad fuel rods meet the same

fuel assembly and fuel rod design bases as VANTAGE 5H fuel

assemblies. Since the original design criteria are met, the ZIRLO

clad fuel rods will not be an initiator for any new accident. The

clad material is similar in chemical composition and has similar

physical and mechanical properties to Zircaloy. Thus, cladding

integrity is maintained and the structural integrity of the fuel

assembly is not affected. ZIRLO cladding improves corrosion

performance and dimensional stability. No concerns have been

identified with respect to the mixed core of Zircaloy and ZIRLO clad

assemblies. Also, no concerns have been identified with respect to

the use of an individual assembly containing a combination of

Zircaloy and ZIRLO clad fuel rods.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not result in any equipment exceeding

its design limits under normal or accident conditions. All design

and performance criteria continue to be met and no new failure

mechanisms have been identified. The ZIRLO cladding material offers

improved corrosion resistance and structural integrity.

The proposed changes do not affect the operation of any system

or component in the plant. The safety functions of the related

structures, systems, or components are not changed, nor is the

reliability of any structure, system, or component reduced. The

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

and do not change any facility design feature, structure, or system.

No new or different type of equipment will be installed. Since there

is no other change to the facility or operating procedures, and the

safety functions and reliability of structures, systems, or

components are not affected, the proposed changes do not create the

possibility of a new accident or an accident different from those

previously evaluated.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

Use of ZIRLO fuel cladding material will not result in any

equipment exceeding its design or licensing bases limits under

normal or accident conditions. VANTAGE 5H reload design and safety

analysis limits are unchanged. For each cycle reload core, the fuel

assemblies will be evaluated using NRC-approved reload design

methods, including consideration of the core physics analysis

peaking factors and core average linear heat rate effects. ZIRLO

fuel assemblies will be assessed for use under conditions consistent

with normal core operating conditions allowed in the Technical

Specifications. 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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869

NRC Project Director: William D. Beckner

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

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

Michigan

Date of amendment request: March 26, 1997

Description of amendment request: The proposed amendment would

modify the technical specifications (TSs) which describe the control

room ventilation system autostart functions.

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:

Per 10 CFR 50.92, the proposed changes do not involve a

significant hazards consideration if the proposed changes do not:

1. involve a significant increase in the probability or

consequences of an accident previously evaluated;

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.

Criterion 1

These changes are administrative in nature, intended to correct

and clarify the TS description of control room ventilation system

operation. Because no changes to plant operations or physical

changes to the plant will occur due to these changes, they do not

involve a significant increase in the probability or consequences of

a previously evaluated accident.

Criterion 2

Because no changes to plant operations or the physical plant

will occur due to these changes, the changes will not create the

possibility of a new or different kind of accident from any

previously evaluated.

Criterion 3

These changes are administrative in nature, intended to correct

and clarify the present TSs with regard to system operation

descriptions. Thus, the changes involve no reduction in 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: Maud Preston Palenske

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

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

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

NRC Project Director: Gail H. Marcus

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

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

Michigan

Date of amendment request: March 26, 1997

Description of amendment request: The proposed amendment would make

three administrative changes to the technical specifications (TSs)

dealing with a grammatical error, an inadvertently deleted frequency

requirement, and a footnote which is no longer applicable.

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:

Per 10 CFR 50.92, the proposed changes do not involve a

significant hazards consideration if the proposed changes do not:

1. involve a significant increase in the probability or

consequences of an accident previously evaluated;

[[Page 27797]]

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.

Criterion 1

This amendment request does not involve a significant increase

in the probability or consequences of an accident previously

evaluated because the proposed changes to the TSs do not affect the

assumptions, parameters, or results of any UFSAR accident analysis.

The firstproposed change, ``A'', is a grammatical correction; the

second proposed change, ``B'', reformats the page, and returns a

frequency requirement that, while inadvertently deleted from the

TSs, was still met via procedure; the third proposed change deletes

a footnote which is no longer applicable. As described in Section

II.C. of licensee's application request dated March 26, 1997, a load

drop analysis is not required for single-failure-proof load blocks.

Criterion 2

The proposed changes do not involve physical changes to the

plant or changes in plant operating configuration. The changes

described above are essentially administrative in nature, and thus

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

from any accident previously evaluated.

Criterion 3

The proposed changes are essentially administrative in nature.

Per NUREG-0612, single-failure-proof cranes are exempt from the

requirements of a load drop analysis; therefore, there is no

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: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Gail H. Marcus

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

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

Date of amendment request: February 12, 1997

Description of amendment request: The proposed amendment would

change position titles in certain Seabrook Station, Unit No. 1

(Seabrook) Appendix A Technical Specifications (TS) to reflect the

present Seabrook organization, would clarify the approval authority for

the Station Qualified Reviewer Program, and would correct a reference.

Specifically, the proposed amendment would:

1. Change TS 6.0, ``Administrative Controls'' to reflect accurately

the current North Atlantic Management organization, their assigned

duties as previously reported to the NRC, and their proper titles,

2. Corrects an incorrect reference in TS 6.4.3.9.b., and

3. Clarifies the term ``Manager'' in TS 6.4.2, ``Station Qualified

Reviewer Program.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below.

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

probabilityor consequences of an accident previously evaluated (10

CFR 50.92(c)(1)) because the proposed changes are merely

administrative or editorial in nature. The proposed changes involve

position title changes to reflect current organization, correct an

incorrect reference, and provide clarification with regard to the

organizational level for certain approvals. The changes do not

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

change any facility design feature or equipment. Since there is no

change to the facility or operating procedures, there is no effect

upon the probability or consequences of any accident previously

analyzed.

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

different kind of accident from any accident previously evaluated

(10 CFR 50.92(c)(2)) because they do not affect the manner by which

the facility is operated or involve any changes to equipment or

features which affect the operational characteristics of the

facility. Therefore, no new accident initiator is introduced that

could cause a new or different kind of accident from those

previously evaluated. The proposed changes merely involve position

title changes to reflect current organization, correct an incorrect

reference, and provide clarification with regard to the

organizational level for certain approvals.

C. The changes do not involve a significant reduction in a

margin of safety (10 CFR 50.92(c)(3)) because the proposed changes

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

involve equipment or features which affect the operational

characteristics of the facility.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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast

Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270.

NRC Project Director: Patrick D. Milano

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-

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

County, Connecticut

Date of amendment request: April 15, 1997

Description of amendment request: The proposed amendment would make

changes to Technical Specification Sections 4.3.3.6 and 4.6.4.1, which

require that the hydrogen monitors be periodically tested.

Specifically, the changes to the surveillances would increase the

testing of the monitor's hydrogen sensor, correct inconsistencies

between surveillances, and make changes to the Bases of the

surveillances.

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:

NNECO [Northeast Nuclear Energy Company] has reviewed the

proposed changes in accordance with 10CFR 50.92 and has concluded

that the change does not involve a significant hazards consideration

(SHC). The bases for this conclusion is that the three criteria of

10CFR 50.92(c) are not satisfied. The proposed changes do not

involve [an] SHC because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to Technical Specification Surveillances

4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test

once per 92 days on a staggered test basis is consistent with the

design and operation of the hydrogen monitor system. The hydrogen

monitoring system is independent of the reactor coolant system

boundary, has no effect on the probability of occurrence of a loss

of coolant accident and performing surveillance testing does not

significantly increase the probability of an accident previously

evaluated.

The proposed change to Technical Specification Surveillances

4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test

will not require the opening of a containment isolation valve and

conducting surveillance testing does not significantly increase the

consequence of an accident previously evaluated.

The proposed change to Technical Specification Surveillances

4.3.3.6 and 4.6.4.1 to change the channel check frequency from once

per 31 days to once per 12 hours on Table 4.3-7 Item 18, add an

analog channel operational test to surveillance 4.3.3.6.2 and make

editorial changes to the surveillances and bases sections are

considered administrative changes. Administrative changes do not

involve a significant increase in the

[[Page 27798]]

probability or consequence of an accident previously evaluated.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes to Technical Specification Surveillances

4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test do

not add any new equipment to the plant and do not affect the way any

system important to safety is operated either in normal or under

accident conditions.

The proposed changes to Technical Specification Surveillances

4.3.3.6 and 4.6.4.1 to change the channel check frequency from once

per 31 days to once per 12 hours on Table 4.3-7 Item 18, add an

analog channel operational test to surveillance 4.3.3.6.2 and make

editorial changes to the surveillances and bases sections are

considered administrative changes.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

The proposed changes to Technical Specification Surveillances

4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test

will provide assurance of expected instrument performance under

accident conditions and performing surveillance testing do not

involve a significant reduction in a margin of safety.

The proposed changes to Technical Specification Surveillances

4.3.3.6 and 4.6.4.1 to change the channel check frequency from once

per 31 days to once per 12 hours on Table 4.3-7 Item 18, add an

analog channel operational test to surveillance 4.3.3.6.2 and make

editorial changes to the surveillances and bases sections are

considered administrative changes. Administrative changes do not

involve a significant reduction in a margin of safety.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed changes do not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

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

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

CT 06141-0270NRC Deputy Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-

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

County, Connecticut

Date of amendment request: April 17, 1997

Description of amendment request: The proposed amendment would

modify Technical Specification 3.7.14 by clarifying the actions to be

taken when an area temperature exceeds its temperature limit.

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:

NNECO [Northeast Nuclear Energy Company] has reviewed the

proposed change in accordance with 10CFR 50.92 and has concluded

that the change does not involve a significant hazards consideration

(SHC). The bases for this conclusion is that the three criteria of

10CFR 50.92(c) are not satisfied. The proposed change does not

involve [an] SHC because the change would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed change to Technical Specification 3.7.14 will

establish allowable tolerances to ensure that the applicable

systems, structures and components are operated within their

existing design bases.

Technical Specification 3.7.14 specifies the actions to be taken

when an area temperature exceeds its temperature limit. The action

taken is dependent on the amount and duration by which the area

temperature exceeds its limit. Actions are currently specified for

exceeding area temperature by less than 20 deg.F and greater than

20 deg.F for periods less than 8 hours and for periods greater than

8 hours. This change clarifies the actions to be taken when the

temperature exceeds its limit by exactly 20 deg.F or exceed its

limit for exactly 8 hours. It is concluded that this change is a

clarification only in that it causes the more conservative actions

to be taken at greater than or equal to 20 deg.F, or at greater

than or equal to 8 hours.

The proposed change, therefore, does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

Establishment of tolerances and clarification of actions at a

specific value does not [ ] change the operation of any system,

structure or component during normal or accident conditions.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

The change is administrative in nature in that it resolves a

discontinuity in the range of temperatures and in the duration

period above the applicable limit for which action is required.

Establishment of tolerances ensures parameters are set and

maintained within allowable design constraints. Clarification of

applicability for the required actions ensures that action is

proscribed for all possible conditions thereby not permitting

operation outside of allowable design.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed change does not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

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

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

CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 31, 1997

Description of amendment request: The proposed amendment would

change Technical Specification (TS) Sections 3/4.6.5.3.2, ``Filtration,

Recirculation, and Ventilation System (FRVS),'' to (1) provide an

appropriate Limiting Condition for Operation and ACTION Statement that

reflects the design basis for the FRVS, and (2) clarify the manner in

which FRVS testing is performed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or

[[Page 27799]]

consequences of an accident previously evaluated.

The proposed TS revisions involve: 1) no hardware changes; 2) no

significant changes to the operation of any systems or components in

normal or accident operating conditions; and 3) no changes to

existing structures, systems or components. Therefore these changes

will not increase the probability of an accident previously

evaluated. Since the plant systems associated with these proposed

changes will still be capable of: 1) meeting all applicable design

basis requirements; and 2) retaining the capability to mitigate the

consequences of accidents described in the HC [Hope Creek] UFSAR

[Updated Final Safety Analysis Report], the proposed changes were

determined to be justified. As a result, these changes will not

involve a significant increase in the 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 changes contained in this submittal will not

adversely impact the operation of any safety related component or

equipment. Since the proposed changes involve: 1) no hardware

changes; 2) no significant changes to the operation of any systems

or components; and 3) no changes to existing structures, systems or

components, there can be no impact on the potential occurrence of

any accident. Furthermore, there is no change in plant testing

proposed in this change request which could initiate an event.

Therefore, these changes will 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 proposed changes for the TS related to the Filtration

Recirculation and Ventilation System (FRVS) Recirculation Subsystem

provide consistency between the Hope Creek TS and post-accident

descriptions of the FRVS Recirculation Subsystem operation already

contained in the UFSAR and reflected in the Hope Creek SER [Safety

Evaluation Report] (NUREG-1048). PSE&G [Public Service Electric &

Gas] believes that the proposed allowed outage times and ACTION

Statements for the FRVS Recirculation Subsystem: 1) will ensure that

the required minimum number of FRVS recirculation units will be

available to mitigate the consequences of accidents described in the

UFSAR; and 2) provide appropriate direction and time requirements

for placing the unit in a safe shutdown condition when the system is

degraded. Therefore, the changes contained in this request do not

result in a significant reduction in a margin of safety.

The revisions to Surveillance Requirement 4.6.5.3.2.b provide an

accurate and clearly defined basis for performing this surveillance

test. The proposed changes implement PSE&Gs existing interpretation

of the TS requirements and therefore do not alter the manner in

which this surveillance test is currently being performed. PSE&G has

concluded that this surveillance test method appropriately tests the

FRVS Recirculation Subsystem. Since the FRVS recirculation units

will continue to be tested with the heaters: 1) operable; and 2) set

at the demand necessary to ``reduce the buildup of moisture,'' PSE&G

believes that the proposed changes to clarify the TS are justified.

Therefore, the changes contained in this request do not result in 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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

Attorney for licensee: M. 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 No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 31, 1997

Description of amendment request: The proposed amendment would

provide changes to Technical Specification (TS) 2.1.2, ``THERMAL POWER,

High Pressure and High Flow,'' ACTION a.1.c for TS 3.4.1.1,

``Recirculation Loops,'' and the Bases for TS 2.1, ``Safety Limits.''

These changes are being made to implement an appropriately conservative

Safety Limit Minimum Critical Power Ratio (SLMCPR) for all Hope Creek

core and fuel designs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The derivation of the revised SLMCPRs for Hope Creek for

incorporation into the Technical Specifications, and its use to

determine cyclespecific thermal limits, have been performed using

NRC approved methods. Additionally, interim implementing procedures

which incorporate cyclespecific parameters have been used which

result in a more restrictive value for SLMCPR. These calculations do

not change the method of operating the plant and have no effect on

the probability of an accident initiating event or transient.

There are no significant increases in the consequences of an

accident previously evaluated. The basis of the MCPR [Minimum

Critical Power Ratio] Safety Limit is to ensure that no mechanistic

fuel damage is calculated to occur if the limit is not violated. The

new SLMCPRs preserve the existing margin to transition boiling and

the probability of fuel damage is not increased. Therefore, the

proposed change does not involve an 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 changes contained in this submittal result from an

analysis of the Cycle 7 core reload using the same fuel types as

previous cycles. These changes do not involve any new method for

operating the facility and do not involve any facility

modifications. No new initiating events or transients result from

these changes. Therefore, the proposed Technical Specification

changes do 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 margin of safety as defined in the Technical Specification

bases will remain the same. The new SLMCPRs are calculated using NRC

approved methods which are in accordance with the current fuel

design and licensing criteria. Additionally, interim implementing

procedures, which incorporate cyclespecific parameters, have been

used. The MCPR Safety Limit remains high enough to ensure that

greater than 99.9% of all fuel rods in the core will avoid

transition boiling if the limit is not violated, thereby preserving

the fuel cladding integrity. Therefore, the proposed Technical

Specification changes do not involve a reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Attorney for licensee: M. 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: April 11, 1997

Description of amendment request: The proposed amendments would

change Technical Specification 3.6.2.3, ``Containment Cooling System''

and the

[[Page 27800]]

associated bases. The changes would increase the cooling water flow

rate for the 31-day and 18-month surveillances and specify that during

the 31-day surveillance the fans are started and operated in low speed.

The changes are being proposed to ensure that the cooling water flow

rate and the fan speed being verified are representative of the

Containment Fan Cooling Unit post-accident mode of operation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes ensure that the fan speed and cooling water

flow rate being verified is representative of the fan speed and

cooling water flow rate required for the post-accident mode of

operation. The proposed changes affect an accident mitigation system

and are being made to assure that the system is being tested in its

accident mitigation mode. There are no new accident initiators

created by the proposed changes. Therefore, the proposed changes do

not involve a significant increase in the probability of an accident

previously evaluated.

The proposed changes provide assurance that the CFCUs will be

capable of maintaining peak containment pressure and temperature

within design limits by verifying the proper post-accident cooling

water flow to the CFCUs. No physical changes to the plant result

from the proposed changes to the surveillance requirements.

Therefore, the proposed changes do not involve a significant

increase in the 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 changes for demonstrating operability of the CFCUs

in the low speed mode, with the required post-accident cooling water

flow rate, are consistent with the existing safety function of the

CFCUs following a Design Basis Accident (DBA). The proposed changes

to the surveillance requirements do not involve any physical changes

to plant components, systems or structures, or the operation of the

CFCUs in the post-accident mode. Therefore, the proposed changes do

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

from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes to the surveillance requirements provide

assurance that the CFCUs will perform their intended design function

of maintaining peak containment pressure and temperature consistent

with the current design basis following a DBA by verifying the

proper post-accident cooling water flow to the CFCUs. Since the high

speed and low speed control circuits are independent and there are

separate breakers used to energize the CFCU motors in high and low

speed, the CFCUs would be capable of starting in the low speed mode

following a DBA although the high speed breaker and control circuit

may not be available.

Verification of the post-accident flow rate during the 31 day

surveillance also ensures that the required supporting system,

Service Water, is available for normal operation. To ensure that the

containment air temperature is maintained below the initial

temperature condition assumed in the accident analysis during normal

operation, Technical Specification 3/4.6.1.5 requires verification

of the average containment temperature once every 24 hours in Modes

1 through 4.

The proposed changes to the CFCU surveillance requirements do

not affect the ability of the CFCUs to perform their normal and

post-accident functions. These proposed changes ensure the

verification of the proper post-accident service water flow rate to

the CFCUs. Therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 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

South Carolina Electric & Gas Company (SCE&G), South Carolina

Public Service Authority, Docket No. 50-395, Virgil C. Summer

Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: March 26, 1997

Description of amendment request: The proposed amendment would

revise the Virgil C. Summer Nuclear Station Technical Specifications to

change the definition of ``Core Alteration.'' The proposed definition

will not consider movement of components other than fuel, sources, or

reactivity control components. These proposed changes are technically

consistent with the requirements of NUREG-1431, Revision 1,

``Westinghouse Standard Technical Specifications,'' issued on April 7,

1995.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes revise the definition of Core Alteration to be

the movement of fuel, sources, or reactivity control components; and to

delete ``or manipulation'' and ``conservative'' from the text. These

changes do not affect the probability of an accident previously

evaluated. The movement of components other than fuel, sources, and

reactivity control components, within the reactor vessel is enveloped

by the analyzed event. Deleting the words ``or manipulation'' and

``conservative'' from the definition of Core Alteration are

administrative changes and also do not impact initiators of analyzed

events. The only component assumed to be an initiator of an analyzed

event is dropping an irradiated fuel assembly, however, fuel is still

part of the definition. Furthermore, a fuel handling accident is

minimized by administrative controls and physical limitations imposed

on fuel handling operations. The movement of components other than

fuel, sources, and reactivity control components within the reactor

vessel will be controlled under plant administrative controls. This

change has no effect on the boron dilution event because when boron

concentration is below limits, Core Alterations are restricted to

maintain the maximum Shutdown Margin. Movement of other components will

have a negligible impact on core reactivity.

The changes to the definition of Core Alteration do not increase

the consequences of an accident previously evaluated. The accident

analysis assumes an irradiated fuel assembly is dropped with the

consequences well within the 10 CFR 100 limits. The dropping of

other components was not addressed in the plant safety analyses,

however, the analysis of the dropped fuel assembly encompasses other

components. The consequences of a boron dilution event are not

addressed because Core Alterations are not allowed when the boron

concentration is below limits. These changes do not affect the

mitigation capabilities of any component or system nor do they

affect the assumptions relative to the mitigation of accidents or

transients. Therefore, the change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes revise the definition of Core Alteration to be

the movement of fuel, sources, or reactivity control components; and to

delete ``or manipulation'' and ``conservative'' from the text. The

change does not involve a

[[Page 27801]]

significant change in the design or operation of the plant. The changes

do not involve a physical alteration of the plant (no new or different

type of equipment will be installed), or new or unusual operator

actions. The changes will not impose any new or different requirements

or eliminate any existing requirements. The definition of Core

Alteration is being clarified and made consistent with NUREG-1431, Rev.

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

different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in margin of

safety?

The proposed changes revise the definition of Core Alteration to

be the movement of fuel, sources, or reactivity control components;

and to delete ``or manipulation'' and ``conservative'' from the

text. The safety analysis assumes an irradiated fuel assembly is

dropped. Controls for handling components other than fuel, sources,

or reactivity control components within the reactor vessel are in

plant administrative controls. The effect of a boron dilution event

on Shutdown Margin is limited due to the requirement to suspend Core

Alterations. The movement of other components have a negligible

impact on core reactivity. No change is being proposed, in the

applicability of the definition, to the movement of components which

factor in the design basis analyses (fuel handling accident).

Deleting the terms ``or manipulation'' and ``conservative'' from the

definition of Core Alteration results in a clarification to the

definition that does not technically alter the meaning. Therefore,

the 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: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180

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

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

NRC Project Director: F. Mark Reinhart, Acting

South Carolina Electric & Gas Company (SCE&G), South Carolina

Public Service Authority, Docket No. 50-395, Virgil C. Summer

Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: March 26, 1997

Description of amendment request: The proposed amendment would

revise the Virgil C. Summer Nuclear Station Technical Specifications

(TS), Surveillance Requirement (SR) 4.5.2.a, to add (1) the charging/

high head safety injection (HHSI) pump cross connect valves, and (2)

the charging pump mini-flow header isolation valve, to the SR valve

list. The proposed change is an administrative change to meet the

recommendations of NRC Branch Technical Position (BTP) EICSB 18, which

establishes the acceptability of disconnecting power to electrical

components of fluid systems as one means of designing against a single

failure that might cause an undesirable component action. TS SR 4.5.2.a

includes a list of the required positions of manually-controlled,

electrically-operated valves, and identify those valves to which the

requirements for removal of electrical power is applied in order to

satisfy the single failure criterion.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change adds the charging/HHSI pump cross connect

valves and the charging pump mini-flow header isolation valve to the

ECCS [Emergency Core Cooling System] Subsystems - Tavg

(greater than or equal to) 350 deg.F Technical Specification

Surveillance Requirement. This Surveillance Requirement will require

the valves to be verified open with power to the valve operators

removed once per 12 hours. ... The charging/HHSI pump cross connect

valves and the charging mini-flow header isolation valve are not

initiators of any analyzed event. ... The charging pump/HHSI pump

cross connect valves are being modified to meet the recommendations

of the BTP (including this Technical Specification change). The

charging pump mini-flow header isolation valve meets the

requirements of the BTP except it is not located in the Technical

Specifications. ... Requiring the valves to be verified open with

power removed from the valve operator once per 12 hours does not

affect the assumptions relative to the mitigation of accidents or

transients. This requirement ensures that the valves are in a

position with power removed so that a failure will not occur that

will affect the mitigation of an accident. These valves are required

to be open during a LOCA [loss-of-coolant accident]. This change

will ensure that the valves are open with power removed. Therefore,

the change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

...This change does not involve a significant change in the

design or operation of the plant. This change is a result of BTP

EICSB 18. The charging/HHSI pump cross connect valves are being

modified to have power lockout capability, redundant indication on

the main control board, and be included in the Technical

Specifications. This will ensure that a single failure (hot short in

the controls of either valve) will not cause spurious actuation of

the valves during the injection or recirculation phase of the ECCS.

The charging pump mini-flow header isolation valve meets the

requirements of the BTP except it is not located in the Technical

Specifications. The charging/HHSI pump cross connect valves and

charging pump mini-flow header isolation valve are required to

remain open during a LOCA. This modification will ensure that the

valves will remain open during an accident which requires ECCS

operation. The proposed change will not introduce any new accident

initiators. Therefore, the change does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

safety?

...The ECCS is required to operate upon receipt of a safety

injection signal. The charging/HHSI pump cross connect valves and

the charging pump mini-flow header isolation valve are required to

remain open during ECCS operation. However, a single failure may

cause a spurious actuation (closure) of the valves which could

hinder HHSI flow. The modification to the charging/HHSI cross

connect valves (the addition of a power lockout feature and

redundant position indication) and the added TS Surveillance

Requirement will eliminate this failure scenario and ensure the

valves remain in their safety function position (open). The charging

pump mini-flow header isolation valves already contain a power

lockout feature and redundant position indication. These valves are

being added to the Technical Specifications to meet the requirements

of BTP EICSB 18. Therefore, the 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: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180

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

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

NRC Project Director: Mark Reinhart, Acting

[[Page 27802]]

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

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

Date of application for amendments: March 13, 1997 (TS 97-01)

Brief description of amendments: The amendments change the

Technical Specifications by raising the allowable U-235 enrichment, as

specified in Section 5.6.1.2, of fuel stored in the new fuel pit

storage racks from 4.5 to 5.0 weight percent.

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:

Operation of Sequoyah Nuclear Plant (SQN) in accordance with the

proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to the allowed enrichment of new fuel stored

in the new fuel storage racks does not change the criticality

potential with the proposed fuel arrangement requirements for the

storage racks. The potential keff values are maintained

the same as the current TS [Technical Specification] requirements.

In addition, the storage racks are not modified, other than the

locations that cannot be filled with fuel assemblies, and the

processes for loading and unloading fuel in these racks and the

controls for these racks remain the same. Since the keff

limits and operating processes are unchanged by the proposed

revision, there is no increase in the probability of an accident

previously evaluated. Likewise, there is no impact to the

consequences of an accident or increase in offsite dose limits as a

result of the proposed TS change because the criticality

requirements are unchanged and plant equipment will be utilized and

operated without change considering the fuel storage location limits

imposed by this request.

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

from any previously analyzed.

As stated above, the plant equipment and operating processes

will not be altered by the proposed TS change with the exception of

allowed fuel storage locations in the new fuel storage racks. The

limitations on acceptable fuel storage locations in the racks ensure

that the keff limits are maintained at the same limits as

currently required. TVA has not postulated a criticality event at

SQN for the spent or new fuel storage locations because the design

of the associated storage racks, potential moderation, and TS

allowable fuel enrichments do not support the potential for this

condition. Considering the physical barriers that will be installed

and verified to be in place prior to initial loading of fuel in the

new fuel storage racks, the new fuel storage rack physical

limitations will continue to ensure that criticality events are not

credible for the proposed change. Therefore, this change does not

create the potential for a new accident from any previously

analyzed.

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

The proposed TS change maintains the existing requirements for

criticality by utilizing limited storage locations in the new fuel

pit storage racks. There is no change to operating practices

associated with the use and control of these racks except for the

storage limitations. For these reasons, there will be no reduction

in the margin of safety as a result of implementing the proposed TS

change.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Creeks, Manitowoc County, Wisconsin

Date of amendment request: April 4, 1997 (TSCR 197)

Description of amendment request: The proposed amendments revise TS

15.6, ``Administrative Controls,'' and 15.7, ``Radiological Effluent

Technical Specifications,'' to change the corporate officer responsible

for nuclear operations from ``Vice President-Nuclear Power,'' to

``Chief Nuclear Officer,'' and to require that the position be an

officer of the company.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not result in a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes are administrative only. There are no

physical changes to the facility or its operation. All Limiting

Conditions of Operation, Limiting Safety System Settings, and Safety

Limits specified in the Technical Specification remain unchanged.

Additionally, there are no changes in the Quality Assurance Program,

Emergency Plan, Security Plan, and Operator Training and

Requalification Program. Therefore, an increase in the probability

or consequences of an accident previously evaluated cannot occur.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes are administrative only. No changes to the

facility structures, systems and components or their operation will

result. The design and design basis of the facility remain

unchanged. The plant safety analyses remain current and accurate. No

new or different failure mechanisms are introduced. Therefore, the

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

previously evaluated is not introduced.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not involve a significant reduction in

a margin of safety.

The proposed amendments are administrative only. All safety

margins established through the design and facility license

including the Technical Specifications remain unchanged. In

addition, the proposed amendments ensure continued emphasis and

assignment of responsibility for overall nuclear safety. Therefore,

all margins of safety are maintained.

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

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

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

amendment request involves no significant hazards consideration.

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

Sixteenth Street, Two Rivers, Wisconsin 54241

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

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

NRC Project Director: John N. Hannon

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

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

Creeks, Manitowoc County, Wisconsin

Date of amendment request: April 14, 1997 (TSCR 198)

Description of amendment request: The proposed amendments revise TS

15.3.1, ``Reactor Coolant System,'' to require both reactor coolant

pumps to be operable when the reactor is critical and to require that

the reactor be placed in hot shutdown within 6 hours if one or both

reactor coolant pumps cease operating. This revision eliminates the

current provision which allows single pump operation up to 3.5 percent

power.

Basis for proposed no significant hazards consideration

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

[[Page 27803]]

licensee has provided its analysis of the issue of no significant

hazards consideration which is presented below:

1. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not result in a significant increase in

the probability or consequences of an accident previously evaluated.

The amendments proposed eliminate an inconsistency in the

Technical Specifications in a conservative manner. The proposed

changes ensure that required protection functions remain operable in

all required modes of operation. Since the protection functions

remain operable in accordance with existing Technical Specification

requirements and serve to mitigate analyzed events no increase in

the consequences of a previously analyzed accident results. The

protective functions are not accident initiators and are maintained

and tested in accordance with existing Technical Specification

requirements, therefore the probability of a previously analyzed

accident cannot increase. Therefore, operation of the Point Beach

Nuclear Plant in accordance with the proposed changes does not

result in an increase in probability or consequences of a previously

analyzed accident.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed amendments restore consistency within the Technical

Specifications thus ensuring the protections functions remain

operable as required and the units are operated within the bounds of

the existing safety analyses. Therefore, operation of the Point

Beach Nuclear Plant in accordance with the proposed amendments does

not result in a new or different kind of accident from any accident

previously evaluated.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not involve a significant reduction in

a margin of safety.

Margins of safety are defined by the bounds of the design and in

the safety analyses performed for the Point Beach Nuclear Plant. The

proposed amendments eliminate an inconsistency within the Technical

Specifications and ensure the plant will respond as analyzed in the

Safety Analyses. There is no physical change in the facility or

operation. Therefore, operation of the Point Beach Nuclear Plant in

accordance with the proposed amendments does not involve a reduction

in 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

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

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

NRC Project Director: John N. Hannon

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: March 21, 1997, as supplemented by

letter dated April 15, 1997.

Description of amendment request: This amendment request proposes

to revise the technical specifications associated with the inspection

of the reactor coolant flywheel to provide an exception to the

recommendations of Regulatory Guide 1.14, Revision 1, ``Reactor Coolant

Pump Flywheel Integrity.'' The proposed exception would allow either an

ultrasonic volumetric examination or surface examination to be

performed at approximately 10-year intervals. In addition, a correction

of the issuance date of a referenced regulatory guide is included.

This amendment would also allow delaying the complete flywheel

examination for the ``D'' reactor coolant pump until the Fall 1997

outage.

This supersedes the staff's proposed no significant hazards

consideration determination evaluation for the requested changes that

was published on January 2, 1997 (62 FR 133).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The safety function of the RCP [reactor coolant pump] flywheels

is to provide a coastdown period during which the RCPs would

continue to provide reactor coolant flow to the reactor after loss

of power to the RCPs. The maximum loading on the RCP flywheel

results from overspeed following a LOCA [loss-of-coolant accident].

The maximum obtainable speed in the event of a LOCA was predicted to

be less than 1500 rpm. Therefore, a peak LOCA speed of 1500 rpm is

used in the evaluation of RCP flywheel integrity in WCAP-14535. This

integrity evaluation shows a very high flaw tolerance for the

flywheels. The proposed change does not affect that evaluation.

Reduced coastdown times due to a single failed flywheel is bounded

by the locked rotor analysis, therefore, it would not place the

plant in an unanalyzed condition. Therefore, these changes do 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 amendment does not create the possibility of a new

or different kind of accident from any previously evaluated since

the proposed amendments will not change the physical plant or the

modes of plant operation defined in the facility operating license.

No new failure mode is introduced due to the proposed change, since

the proposed change does not involve the addition or modification of

equipment, nor do they alter the design or operation of affected

plant systems, structures, or components.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The operating limits and functional capabilities of the affected

systems, structures, and components are basically unchanged by the

proposed amendment. The results of the flywheel inspections

performed have identified no indications affecting flywheel

integrity. As identified in WCAP-14535, detailed stress analysis as

well as risk analysis have been completed with the results

indicating that there would be no change in the probability of

failure for RCP flywheels if all inspections were eliminated.

Therefore these changes do not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

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

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

66621

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

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037

NRC Project Director: William H. Bateman

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

[[Page 27804]]

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. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, located in Grundy County,

Illinois

Date of amendment request: January 24, 1997.

Description of amendment request: The application proposed to

change the Technical Specifications to reflect the installation of new

reactor water level instrumentation for the Emergency Core Cooling

System actuation.

Date of publication of individual notice in Federal Register: April

18, 1997 (62 FR 19143).Expiration date of individual notice: May 19,

1997

Local Public Document Room location: The Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450

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

Nuclear Power Station, Units 2 and 3, located in Grundy County,

Illinois

Date of amendment request: March 5, 1997.

Description of amendment request: The application proposed to

remove the Main Steam Line Radiation Monitor High scram and the Main

Steam Line Tunnel Radiation High input to the Main Steam Line Isolation

function requirement from the Technical Specifications (TS). The

proposed changes are a result of a Boiling Water Reactor Owners Group

initiative to minimize inadvertent scrams and Main Steam Isolation

Valve closure due to erroneous radiation monitor actuation.

Date of publication of individual notice in Federal Register: April

18, 1997 (62 FR 19141).Expiration date of individual notice: May 19,

1997

Local Public Document Room location: The Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad

Cities Nuclear Power Station, Units 1 and 2, located in Rock Island

County, Illinois

Date of amendment request: April 21, 1997

Description of amendment request: The amendments would reflect a

change in the Quad Cities, Unit 2, Minimum Critical Power Ratio (MCPR)

Safety Limit and add the Siemens Power Corporation (SPC) methodology

for application of the Advanced Nuclear Fuel for Boiling Water Reactors

(ANFB) Critical Power Correlation to coresident General Electric fuel

for Quad Cities, Unit 2, Cycle 15, to Technical Specification Section

6.9.A.6.b.

Date of publication of individual notice in Federal Register: April

30, 1997 (62 FR 23499)

Expiration date of individual notice: May 30, 1997

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

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

Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: March 31, 1997

Brief description of amendment: The proposed amendment would revise

the Ginna Station Improved Technical Specifications to reflect a

planned modification to the spent fuel pool storage racks.Date of

publication of individual notice in Federal Register: April 30, 1997

(62 FR 23502)

Expiration date of individual notice: May 30, 1997

Local Public Document Room location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610

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.

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 application for amendments: March 5, 1997, as supplemented

May 9, 1997. The May 9, 1997, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination.

Brief description of amendments: The amendments incorporate a new

Technical Specification for instrumentation associated with automatic

isolation of a pathway for release of non-condensible gases from the

main condenser.

Date of issuance: May 9, 1997

Effective date: May 9, 1997

Amendment Nos.: 185 and 216

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

change the Technical Specifications.

Date of initial notice in Federal Register: April 9, 1997 (62 FR

17224) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 9, 1997.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

North Carolina

Date of application for amendment: March 14, 1997

Brief description of amendment: The amendment extends the allowed

outage time for its refueling water storage tank

[[Page 27805]]

while performing surveillance testing of its reactor coolant system

pressure isolation valves (Surveillance 4.4.6.2.2).

Date of issuance: May 6, 1997

Effective date: May 6, 1997

Amendment No. 71

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

Technical Specifications.

Date of initial notice in Federal Register: March 26, 1997 (62 FR

14459) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 6, 1997.No significant

hazards consideration comments received: No

Local Public Document Room location: Cameron Village Regional

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

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

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

North Carolina

Date of application for amendment: April 18, 1997, as supplemented

April 29, 1997.

Brief description of amendment: The amendment approves the

modification to the protection circuitry for emergency diesel

generators. The associated Safety Evaluation delineates the staff's

review and findings that the modification and related Final Safety

Analysis Report (FSAR) changes are acceptable.

Date of issuance: May 8, 1997

Effective date: May 8, 1997

Amendment No. 72

Facility Operating License No. NPF-63. The amendment approves

modification to the protection circuitry for emergency diesel

generators and related FSAR changes.

Date of initial notice and proposed no significant hazards

consideration in Federal Register: (62 FR 19818 dated April 23, 1997).

The 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 May 23, 1997, 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 determination of

no significant hazards consideration is contained in a Safety

Evaluation dated

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

Local Public Document Room location: Cameron Village Regional

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

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

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

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

Will County, Illinois

Date of application for amendments: November 4, 1996, as

supplemented on December 4, 1996, and March 20, 1997.

Brief description of amendments: The amendments revise the

technical specifications (TS) to permit the removal of containment

tendon sheathing filler grease in up to 35 tendons for Byron, Unit 1,

and Braidwood, Unit 1, in advance of the steam generator replacement

outages. The grease will be removed approximately 6 months prior to the

respective steam generator replacement outages. In addition, in

Amendment No. 80 issued on April 16, 1997, the title in Braidwood's TS

6.9.1.7 was unintentionally left uncorrected. The corrected page is

included in this amendment.

Date of issuance: May 6, 1997

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 89, 89 and 81, 81

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

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 15, 1997 (62 FR

2186). The March 20, 1997, submittal provided additional clarifying

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 May 6, 1997No 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.

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

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

Date of application for amendments: February 17, 1997, as

supplemented February 27, March 12, March 26, April 2, and April 10,

1997

Brief description of amendments: The amendments evaluate the

Unreviewed Safety Question (USQ) associated with the use of containment

pressure to compensate for the deficiency in Net Positive Suction Head

(NPSH) for the Emergency Core Cooling System (ECCS) pumps following a

Design Basis Accident (DBA). In the resolution of the USQ, the licensee

changed the Updated Final Safety Analysis Report (UFSAR) in the

following areas:

1. containment analysis,

2. decay heat model,

3. increase in the suppression pool temperature and the effect on

other associated systems following a DBA, and

4. ECCS heat exchanger duty and containment cooling service water

(CCSW) system flow.In addition, the proposed amendments would change

the Technical Specification (TS) allowable water temperature limits for

the suppression chamber and the ultimate heat sink from less than or

equal to 75 degrees Fahrenheit to less than or equal to 95 degrees

Fahrenheit. The original licensing basis water temperature for both the

suppression chamber and ultimate heat sink was 95 degrees Fahrenheit.

Both values were changed in the TS in Amendment Nos. 152 and 147 for

Dresden, Units 2 and 3, respectively, issued on January 28, 1997. The

amendments to lower the ultimate heat sink and suppression pool

temperature limits in the TS was in response to the resolution of a USQ

associated with the operation of Dresden, Units 2 and 3, following the

discovery of a calculational error concerning the head loss across the

ECCS suction strainers. The proposed amendments will return both units

to normal operating conditions allowing for continued power operations

when the ultimate heat sink temperature goes above 75 degrees

Fahrenheit during warm weather.

Date of issuance: April 30, 1997

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 157; 152.

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

revised the licenses, TS and USFAR.

Date of initial notice in Federal Register: February 27, 1997 (62

FR 8998). The February 27, March 12, March 26, April 2 and April 10,

1997, submittals provided additional clarifying 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

[[Page 27806]]

April 30, 1997No significant hazards consideration comments received:

No

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad

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

Illinois

Date of application for amendments: February 17, 1997

Brief description of amendments: The amendments would change the

Technical Specifications by increasing the load test values of the

emergency diesel generators in Surveillance Requirement 4.9.A.8.h from

between 2625 kW and 2750 kW to 2730 kW and 2860 kW.

Date of issuance: May 1, 1997

Effective date: Immediately, to be implemented within 60 days.

Amendment Nos.: 176 and 172

Facility Operating License Nos. DPR-29 and DPR-30: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: March 26, 1997 (62 FR

14460). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 1, 1997.No significant

hazards consideration comments received: No

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

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

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

Date of application for amendment: August 22, 1996, as supplemented

March 28, 1997.

Brief description of amendment: The amendment revises Technical

Specification Sections 3.3 and 4.5 to allow the deletion of the

requirement to utilize sodium hydroxide (NaOH) as an additive in the

post-accident containment spray system.

Date of issuance: April 23, 1997

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

30 days.

Amendment No.: 191

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

Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4345) The March 28, 1997, supplemental letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination or expand the scope of the

amendment request as originally noticed. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

April 23, 1997No significant hazards consideration comments received:

No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

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

Date of application for amendments: March 7, 1997, as supplemented

by letters dated April 2, 10, 16, 22, and 28, 1997

Brief description of amendments: The amendment revise Section 3/

4.7.1.6 of the Technical Specifications to require four instead of

three steam generator pressure operated relief valves operable.

Date of issuance: April 29, 1997

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

30 days. Implementation of the amendments include the incorporation in

the Updated Final Safety Analysis Report (UFSAR) of the changes to the

description of the facility as set forth in the licensee's application

dated March 7, 1997, as supplemented by letters dated April 2, 10, 16,

22, and 28, 1997, as evaluated in the staff's Safety Evaluation dated

April 29, 1997.

Amendment Nos.: 159 and 151

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

revised the Technical Specifications and License Conditions.

Date of initial notice in Federal Register: March 13, 1997 (62 FR

11931) The April 2, 10, 16, 22, and 28, 1997, letters provided

additional and clarifying information that did not change the scope of

the March 7, 1997, application and the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 29, 1997.No significant hazards

consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

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

supplemented by letters dated September 18, 1995, and March 15, April

29, May 16, September 23, and October 28, 1996, and January 16, April

22, and May 2, 1997

Brief description of amendments: The amendments revise the

Technical Specifications related to the replacement of the Westinghouse

Model ``D'' type preheat steam generators with feedring steam

generators designed by Babcock and Wilcox International.

Date of issuance: May 5, 1997

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

30 days for Unit 1; and effective upon replacement of the steam

generators for Unit 2.

Amendment Nos.: 175 and 157

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 8, 1995 (60 FR

56366) The March 15, April 29, May 16, September 23, and October 28,

1996, and January 16, April 22, and May 2, 1997, letters provided

clarifying information that did not change the scope of the September

30, 1994, application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 5, 1997.No significant hazards

consideration comments received: No

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina 28223-0001

Entergy Gulf States, Inc., 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: November 15, 1996

Brief description of amendment: The amemdment revises the technical

specifications to allow the performance of the 24-hour emergency diesel

generator maintenance run while the unit is in either Mode 1 or Mode 2.

Date of issuance: May 5, 1997

Effective date: May 5, 1997

Amendment No.: 94

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 2, 1997 (62 FR

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

in a Safety Evaluation dated May 5, 1997.No significant hazards

consideration comments received. No.

Local Public Document Room location: Government Documents

[[Page 27807]]

Department, Louisiana State University, Baton Rouge, LA 70803

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

Unit No. 1, Pope County, Arkansas

Date of amendment request: April 11, 1997

Brief description of amendment: The amendment would permit steam

generator tubes with intergranular corrosion indications that may

exceed through-wall limits to remain in service until the next

refueling outage.

Date of issuance: May 7, 1997

Effective date: May 7, 1997

Amendment No.: 189

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

Technical Specifications.Public comments requested as to proposed no

significant hazards consideration (NSHC): Yes (62 FR 19628 dated April

22, 1997). The notice provided an opportunity to submit comments on the

Commission's proposed NSHC determination. No comments have been

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

hearing by May 22, 1997, but indicated that if the Commission makes a

final NSHC 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 determination of

NSHC are contained in a Safety Evaluation dated May 7, 1997.

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

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

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: December 19, 1996

Brief description of amendment: The proposed changes revise

Technical Specification Table 4.3-1 to change the power calibration

requirements for the linear power level, the Core Protection Calculator

(CPC) delta T power and the CPC nuclear power signals between 15 and 80

percent power to allow more conservative settings.

Date of issuance May 5, 1997

Effective date: May 5, 1997, to be implemented within 30 days.

Amendment No.: 183

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

Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4348) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 5, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: November 12, 1996, as

supplemented November 27, 1996 (TSCR 224)

Brief description of amendment: The amendment updates the technical

specifications to reflect the implementation of the revised 10 CFR Part

20, ``Standards for Protection Against Radiation.''

Date of issuance : May 8, 1997

Effective date: May 8, 1997, with full implementation within 30

days.

Amendment No.: 191

Facility Operating License No. DPR-16. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 18, 1996 (61

FR 66708). The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated May 8, 1997.No significant

hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

Atomic Power Station, Lincoln County, Maine

Date of application for amendment: February 7, 1997

Brief description of amendment: The amendment revises TS 3.12,

``Station Service Power,'' to require both 115 kV power circuits to be

operable when the reactor is critical and to limit or restrict the time

during which Maine Yankee may continue to operate if one or both of the

115 kV power circuits become inoperable.

Date of issuance May 2, 1997

Effective date: May 2, 1997, to be implemented within 30 days.

Amendment No.: 157

Facility Operating License No. DPR-36: Amendment revised the

Technical Specifications and/or License.

Date of initial notice in Federal Register: February 26, 1997 (FR

8799) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 2, 1997No significant hazards

consideration comments received: No

Local Public Document Room location: Wiscasset Public Library, High

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

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 February 24, 1997, as supplemented by

letters dated March 13, April 11, 23, and 29, 1997

Brief description of amendments: The amendments change the

Technical Specification surveillance requirements for the Control Room

Emergency Filtration System, the Penetration Room Filtration System,

and the Containment Purge Exhaust Filter System.

Date of issuance May 1, 1997

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

30 days

Amendment Nos.: 127 and 121

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

the Technical Specifications.

Date of initial notice in Federal Register: March 6, 1997 (62 FR

10294) The March 13, April 11, 23, and 29, 1997, letters provided

clarifying information that did not change the scope of the February

24, 1997, application and the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated May 1, 1997.No

significant hazards consideration comments received: No

Local Public Document Room location: Houston-Love Memorial

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

36302

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

Ferry Nuclear Plant, Units 2 and 3, Limestone County, Alabama

Date of application for amendments: June 21, 1996, supplemented

February 7, 1997 (TS 377)

Brief description of amendments: The amendments provide a new

minimum critical power ratio safety limit to replace a nonconservative

value. Technical Specification Bases are also updated to clarify usage

of the residual heat removal system supplemental spent fuel pool

cooling mode.

Date of issuance : May 7, 1997

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

within 30 days from the date of issuance.

Amendment Nos.: 247 and 207

[[Page 27808]]

Facility Operating License Nos. DPR-52 and DPR-68: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register:

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 7, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room location: Athens Public library, 405 E.

South Street, Athens, Alabama 35611

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

50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa

County, Virginia

Date of application for amendments: September 4, 1996, as

supplemented February 3, 1997. The February 3, 1997 submittal provided

clarifying information only, and did not change the proposed no

significant hazards consideration determination.

Brief description of amendments: The amendments revise the license

and technical specifications (TS) to permit the insertion of four

demonstration fuel assemblies into the reactor core of either North

Anna 1 or North Anna 2, as described in the licensee's submittal. The

four lead test assemblies, fabricated by Framatome Cogema Fuels, will

incorporate several advanced design features, including: a debris

filter bottom nozzle, mid-span mixing grids, a floating top end grid, a

quick disconnect top nozzle, and use of advanced zirconium alloys for

fuel assembly structural tubing and for fuel rod cladding.

Date of issuance May 9, 1997

Effective date: May 9, 1997

Amendment Nos.: 204 and 185

Facility Operating License Nos. NPF-4 and NPF-7. These amendments

revised the License and Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64396) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 9, 1997.No significant

hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: July 18, 1996, as supplemented

on January 29, 1997.

Brief description of amendment: The amendment revises Kewaunee

Nuclear Power Plant Technical Specification 3.8, ``Refueling,'' and its

associated Basis, by allowing the containment personnel air lock doors

to remain open during refueling operations.

Date of issuance May 7, 1997

Effective date: May 7, 1997

Amendment No.: 132

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 14, 1996 (61 FR

42285). The January 29, 1997, submittal provided supplemental

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

May 7, 1997.No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

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

Notice Of Issuance Of Amendments To Facility Operating LicensesAnd

Final Determination Of No Significant Hazards ConsiderationAnd

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

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 for the

amendment 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.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective 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

application for amendment, (2) the amendment to

[[Page 27809]]

Facility Operating License, 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 room for the

particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By June 20, 1997, 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. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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 20555-001, 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 20555-001, 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 the

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

Commonwealth Edison Company, Docket No. 50-265, Quad Cities Nuclear

Power Station, Unit 2, Rock Island County, Illinois

Date of application for amendment: April 29, 1997.

Brief description of amendment: The proposed amendment modifies

Section 5.3.A, ``Design Features'' of the Technical Specifications (TS)

to reflect the ATRIUM-9B fuel design and would include various Siemens

Power Corporation (SPC) topical reports in TS Section 6.9.A.6, ``Core

Operating Limits Report,'' to reflect mechanical design criteria for

this fuel and topical reports required for operation. This change would

allow this fuel to be loaded into the core only under Operational Modes

3 (Hot Shutdown), 4 (Cold Shutdown), and 5 (Refueling) and does not

permit startup or power operation using the ATRIUM-9B fuel.

Date of issuance May 2, 1997

Effective date: May 2, 1997

Amendment No.: 173

Facility Operating License No. DPR-30: The amendment revised the

Technical Specifications.Public comments requested as to proposed no

significant hazards consideration: No.The Commission's related

evaluation of the amendment, finding of emergency circumstances and

final determination of no significant hazards consideration are

contained in a Safety Evaluation dated May 2, 1997.

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

Austin, One First National Plaza, Chicago, Illinois 60603

[[Page 27810]]

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

NRC Project Director: Robert A. Capra

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

Station, Nemaha County, Nebraska

Date of amendment request: May 2, 1997, as superseded May 5, 1997.

Brief description of amendment: The proposed amendment relocates

and revises the requirements for the control of the setpoint for the

Standby Liquid Control system relief valves. The requirements would be

relocated from Section 4.4.A.2.a and Bases Section 3.4.A of the Cooper

Technical Specifications to the Updated Safety Analysis Report and the

Inservice Testing Augmented Testing Program.

Date of issuance May 9, 1997

Effective date: May 9, 1997

Amendment No.: 176

Facility Operating License No. DPR-46: The amendment revised the

Technical Specifications.Public comments requested as to proposed no

significant hazards consideration: No.The Commission's related

evaluation of the amendment, finding of emergency circumstances and

final determination of no significant hazards consideration are

contained in a Safety Evaluation dated May 9, 1997.

Local Public Document Room location: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305.

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499

NRC Project Director: William D. Beckner

Dated at Rockville, Maryland, this 14th day of May, 1997.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects III/IV, Office of Reactor

Regulation

[Doc. 97-13190 Filed 5-20-97; 8:45 am]

BILLING CODE 7590-01-F