Biweekly Notice

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 March 29, 1997, through April 11, 1997. The

last biweekly notice was published on April 9, 1997 (62 FR 17223).

Notice of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunith For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the 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 May 23, 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

[[Page 19826]]

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).

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.

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 amendment request: March 17, 1997

Description of amendment request: The proposed change would revise

eight specifications for 18-month tests to delete a conditional

statement that the testing be done while the unit is shut down and to

clarify that Harris Nuclear Plant (HNP) may take credit for tests on

some components which are performed while the unit is at power.

Basis for proposed no significant Hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed changes permit HNP to evaluate the conditions

required to safely perform a test, but the changes do not directly

affect the functioning or operation of any plant equipment. Since no

equipment operation is involved there is no increase in the

probability or consequence of any previously identified accident.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes to the conditional statements on the

surveillance frequencies do not involve any physical alterations or

additions to plant equipment or alter the manner in which any

safety-related system performs itsfunction or is operated.

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

new or different kind of accident from any previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed changes to the conditional statements on the

surveillance frequency allows HNP to evaluate the conditions needed

to safely perform the required testing. There is no change in the

frequency of testing or in the testing which is required. There is

no change in the responsibility of HNP toperform tests in a safe and

responsible manner, and any changes to procedures will have to be

individually evaluated to ensure that they do not reduce the margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

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

Raleigh, North Carolina 27602

NRC Project Director: Mark Reinhart, Acting

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

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

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

Will County, Illinois

Date of amendment request: January 30, 1997

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 1.0, ``Definitions;'' TS 3/4.6.1,

``Primary Containment'' and associated Bases; and TS 5.4.2, ``Reactor

Coolant System Volume'' for Byron and Braidwood to support steam

generator replacement. ComEd will be replacing the original

Westinghouse D4 steam generators at Byron and Braidwood with Babcock

and Wilcox International steam generators. The replacement steam

generators increase the Reactor Coolant System volume which results in

a higher calculated peak containment pressure (Pa) value.

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.

Each of the [replacement steam generators] RSGs has a larger

[reactor coolant system] RCS side volume than the original steam

generators (OSGs). As a result of the RCS

[[Page 19827]]

volume increase, the mass and energy release during the blowdown

phase of the large break loss of coolant accident (LBLOCA) is

increased. Additionally, the heat transfer rate of the RSGs is

greater than the OSGs, and the RSGs will operate at a slightly

higher pressure than that for the OSGs. Consequently, the steam

enthalphy exiting the break during the reflood period, with the RSG,

will be greater than that for the OSG. This results in an increase

in the containment building peak pressure, Pa.

The proposed revisions to the Technical Specifications involve

the specified value of Unit 1 RCS volume and the defined value of

Unit 1 Pa. Several editorial changes are also being made

to improve clarity and consistency of the TS.

RCS volume is not an initiator for any event and an increase in

volume does not affect any operating margin or requirements.

Therefore, increasing the primary volume does not increase the

probability of any event previously analyzed.

The revised value of Pa continues to be less than the

design basis pressure for the containment building structure. The

change represents only a revision to the containment test pressure

for containment leakage testing. Such testing is only performed with

the affected unit in the shutdown condition. Therefore, the proposed

change in Pa does not involve a significant increase in

the probability of an accident previously evaluated.

All accidents in the Updated Final Safety Analysis Report

(UFSAR) were evaluated to determine the effect of an increase in

primary volume on accident consequences. The events identified that

may be impacted by an increase in primary volume are the Waste Gas

System Leak or Failure and LBLOCA. For the Waste Gas System Leak or

Failure, the activity of the decay tank is controlled to Technical

Specification limits which are unaffected by RCS volume. Therefore,

an increase in RCS volume would not increase the offsite dose.

The offsite dose calculation for the LBLOCA is unaffected by the

proposed change. The license basis offsite dose calculation is in

accordance with NRC Reg Guide 1.4 ``Assumptions Used for Evaluating

The Potential Radiological Consequences of a Loss of Coolant

Accident for Pressurized Water Reactors.'' This Regulatory Guide

states, in part, ''...a number of appropriately conservation

assumptions, based on engineering judgment and on applicable

experimental results from safety research programs conducted by the

AEC.'' These conservatisms include (but are not limited to) the

following assumptions:

Twenty five percent of the equilibrium radioactive full

power inventory is immediately available for leakage from the

primary containment.

100% of the equilibrium full power radioactive noble

gas inventory is immediately available for leakage from the primary

containment.

The primary containment should be assumed to leak at

the (maximum) leak rate specified in the technical specifications

for the first 24 hours and at 50% of this value for the remaining 29

days of the accident duration.

The design basis leakage corresponding to a peak containment

pressure of 50 psig utilized in the design basis accident analysis

is 0.10% per day of the containment free air mass. Therefore, the

offsite dose calculation was performed with a leakage of .1% per day

for day one and .05% per day for days two through 30. Isotopic

inventories are unaffected by the increase in reactor coolant

volume. Thus, the offsite dose is unaffected by the increase in the

peak containment pressure. Therefore, this proposed change to

Pa does not involve a significant increase in the

consequences of an accident previously evaluated.

The editorial changes proposed are for clarity and consistency

within the Technical Specifications and do not affect either the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change in RCS volume is a change in a plant

parameter within the ``Design Features'' section of the Technical

Specifications. Increasing the RCS volume does not create any new or

different failure modes. The existing RCS design requirements

continue to be met.

The revised value of Pa continues to be less than the

design basis pressure for the containment building structure. The

change represents only a revision to the test pressure for

containment leakage testing. Such testing is only performed with the

affected unit in the shutdown condition. Therefore, no new or

different failure modes are being introduced by modification of the

testing parameters.

The editorial changes proposed are for clarity and consistency

within the Technical Specifications and do not result in any

physical changes to the facility or how it is operated. No new or

different failure modes are being introduced by these changes.

Therefore, these proposed 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.

Changing the RCS volume in the Technical Specifications does not

reduce the margin of safety. RCS volume is a design feature. The

change in RCS volume does not involve a change to any setpoint or

design requirements. An evaluation of all UFSAR accidents was

performed to determine the effect of an increase in RCS volume. This

evaluation is summarized as follows:

An evaluation of the Chemical and Volume Control System

Malfunction was performed to determine the effect of the increased

RCS volume due to the RSGs. The larger RCS volume of the RSGs

reduces the reactivity insertion for a given dilution flow rate.

Therefore, the UFSAR analyses remain bounding for Byron Unit 1 and

Braidwood Unit 1 with the RSGs and there is no reduction in the

margin of safety.

An evaluation of the Inadvertent Actuation of the Emergency Core

Cooling System During Power Operation Event was performed to

determine the effect of the increased RCS volume due to the RSGs.

For this event, the injection of borated water causes a negative

reactivity insertion, which increases DNBR. For a given Refueling

Water Storage Tank (RWST) boron concentration, the larger RCS volume

will cause a reduction in the negativity insertion rate as compared

to the current UFSAR analysis. However, negative reactivity would

still be inserted, no fuel pins would experience DNB, and there is

no reduction in the margin of safety.

An evaluation of the Small Break LOCA was performed to determine

the effect of increased RCS volume. The additional RCS volume will

cause a delay in the loop seal clearing which in turn delays the

core uncovery as compared with the UFSAR analysis. A delay in core

uncovery reduces the amount of core heatup which results in a lower

peak clad temperature (PCT) because the core decay heat would be

less than in the UFSAR analysis. The benefit is considered small,

but there is still a benefit. Therefore, the increased RCS volume

does not result in a reduction in the margin of safety.

An evaluation of the Large Break LOCA was performed to determine

the effect of increased RCS volume. For a LBLOCA, the increased RCS

volume causes the blowdown phase of the event to be longer.

Increased blowdown phase, alone, could potentially result in a

higher PCT. However, the RSGs also have less resistance to flow due

to increased primary side steam generator flow area, which results

in a higher blowdown flow compared to the OSGs. The increased

blowdown flow more than compensates for the longer blowdown phase

associated with the increased RCS volume. The net effect is a

decrease in PCT for the RSG compared to the OSG. Therefore, there is

no reduction in the margin of safety.

An evaluation of the Gas Waste System Leak or Failure was

performed to determine the effect of the increased RCS volume.

Because the activity of the decay tank is controlled within

Technical Specification limits, an increase in RCS volume would not

change the results of the event. Therefore, there is no reduction in

the margin of safety.

An evaluation was performed to determine the effect of the

increased RCS volume on the peak containment pressure following a

LBLOCA. The increased RCS volume caused the peak containment

pressure to increase to 47.8 psig. This is still below the

containment design pressure of 50.0 psig. Therefore, there is no

reduction in the margin of safety.

This proposed change involves testing requirements designed to

demonstrate adequate leakage rates are maintained. If adequate

leakage rates are maintained as outlined in the Technical

Specifications, there will be no reduction in the margin of safety.

In the event of degradation of a containment seal that results in

unacceptable leakage, plant shutdown will occur as required by

Technical Specifications and administrative requirements in

accordance with approved plant procedures. Therefore, this proposed

change does not involve a significant reduction in a margin of

safety.

The editorial changes proposed are for clarity and consistency

within the Technical Specifications and do not result in any

physical changes to the facility or how it is

[[Page 19828]]

operated. Therefore, the changes have no effect on the margin of

safety.

Thus, this amendment request does not result in any decrease 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

requested amendments involve no significant hazards consideration.

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.

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan

Date of amendment request: March 27, 1997

Description of amendment request: The proposed amendment would

alter the company name in the Facility Operating License DPR-20 and

Technical Specifications for the Palisades Plant. Specifically, the

proposed amendment would revise the name from ``Consumers Power

Company'' to ``Consumers Energy 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:

A. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

Since the proposed changes do not alter the technical content of

any Facility Operating License or Technical Specifications

requirements, they do not alter any feature of plant equipment,

settings, operation, or configuration.

Therefore, they cannot involve a significant increase in the

probability of an accident previously evaluated.

The proposed changes alter the company name in the Facility

Operating License and Technical Specifications to reflect the change

from ``Consumers Power Company'' to ``Consumers Energy Company''.

The proposed change will not affect any obligations. The company

will continue to own all of the same assets, will continue to serve

the same customers, and will continue to honor all existing

obligations and commitments. The proposed changes will not alter

plant operation or configuration, or its ability to respond to

accidents.

Therefore, they will not involve a significant increase in the

consequences of any accident previously evaluated.

B. Do the proposed changes create the possibility of a new or

different kind of accident from any previously evaluated?

Since the proposed changes do not alter the technical content of

any Facility Operating License or Technical Specifications

requirements, they do not alter any feature of plant equipment,

settings, operation or configuration.

Therefore, they cannot create the possibility of a new or

different kind of accident from any previously evaluated.

C. Do the proposed changes involve a significant reduction in a

margin of safety?

Since the proposed changes do not alter the technical content of

any Facility Operating License or Technical Specifications

requirements, they do not alter any feature of plant equipment,

settings, operation, or configuration.

Therefore, they cannot 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: Van Wylen Library, Hope

College, Holland, Michigan 49423

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

Company, 212 West Michigan Avenue, Jackson, Michigan 49201

NRC Project Director: John N. Hannon

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

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

Date of amendment request: March 31, 1997 (TSC 96-10)

Description of amendment request: The proposed amendments would

modify and clarify the High Pressure Injection (HPI) System operability

requirements in Specification 3.3.1, impose additional HPI system

operability requirements for operation above 75 percent power,

incorporate the new Standard Technical Specifications format for the

HPI system, revise Specification 3.3.2 to clarify that the Reactor

Building Emergency Sump isolation valves are remote-manually operated

valves, and add new specifications and a surveillance test to address

operability requirements of the atmospheric dump valves. In addition,

corresponding Bases changes would be incorporated.

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) Involve a significant increase in the probability or

consequences of an accident previously evaluated:

No. None of the proposed changes has any impact upon the

probability of any accident which has been evaluated in the UFSAR

[Updated Final Safety Analysis Report]. The only potential change in

operating configuration is allowing operation with the HPI [High

Pressure Injection] System pump discharge header cross-

connected. This operating mode does not affect the probability

of a LOCA [Loss-of-Coolant Accident] or of any other accident

evaluated in the UFSAR.

None of these changes have any impact upon the ability of the

HPI System to add soluble poison to the Reactor Coolant System, as

addressed by Specification 3.2. The remaining potential impact is

upon the ability to mitigate the consequences of a small break LOCA,

which is addressed below. The small break LOCA is the limiting

design basis accident with respect to HPI System operability

requirements.

The proposed changes to Specification 3.3.1 provide appropriate

actions to address any degradation in the operability of the HPI

System. The operability requirements for the HPI System are

supported by a spectrum of small break LOCA analyses based on the

approved Evaluation Model described in FTI [Framatome Technologies,

Incorporated] topical report BAW-10192P. These small break LOCA

analyses demonstrate that the acceptance criteria of 10CFR 50.46 are

not violated.

Two trains of HPI are required to mitigate a small break LOCA

above 75% FP [full power]. Operability requirements in the proposed

Technical Specifications assure that the HPI System can withstand

the worst single failure and still result in two HPI pumps injecting

through two trains. The full power small break LOCA analyses

supporting this proposed license amendment have been performed in

accordance with the approved Evaluation Model described in FTI

topical report BAW-10192P. The proposed Technical Specifications

limit operation above 75% FP with a degraded HPI System to 72 hours

before a power reduction to less than 75% FP (or a reactor shutdown)

must be initiated. The required actions depend on the HPI System

components that are inoperable. The 72 hour completion time is

consistent with the time requirements for HPI specified in NUREG-

1430.

When at or below 75% FP, one HPI train provides sufficient flow

to mitigate a small break LOCA. The 75% power level is justified by

analyses using the Evaluation Model described in FTI topical report

BAW-10192P, considering the worst case break location and size

described in LER [Licensee Event Report] 269/90-15 and Attachment 3

to this submittal. The proposed Technical Specifications require two

HPI trains to be operable at or below 75% FP. These requirements

ensure that, following the worst single failure, one train of HPI

would remain

[[Page 19829]]

available to mitigate a small break LOCA. Operation with less than

two HPI trains operable is restricted to 72 hours before shutdown

requirements are imposed. This completion time is consistent with

the time requirements specified for an HPI System in NUREG-1430.

The additional HPI system restriction that requires the HPI pump

discharge header to be cross-connected when all three HPI pumps are

operable does not increase the consequences of a small break LOCA.

If a single failure prevents one HPI train from actuating, this

lineup results in at least two HPI pumps initially injecting through

the automatically actuating train. This increases the amount of

cooling flow initially delivered to the core as compared to the

current system configuration.

The impact of this alignment has been evaluated, considering the

potential single active failures, including the failure of any

powered component to operate and any single failure of electrical

equipment.

It has been determined that, when each of the three HPI pumps is

either running or is capable of automatic actuation upon an

Engineered Safeguards signal, cross-connection of the HPI pump

discharge header does not introduce susceptibility to any single

failure. Therefore, the potential consequences of a small break LOCA

are not increased. If fewer than three HPI pumps are either running

or are capable of automatic actuation, and the HPI pump discharge

header were cross-connected, a single failure of one pump could

cause a single pump to be aligned to both HPI trains. In this

condition, the single pump could experience runout conditions prior

to corrective operator action. However, proposed Specification 3.3.1

requires the discharge header to be isolated between the two

remaining operable HPI pumps. The proposed BASES provide guidelines

to ensure that the requirements for redundancy are properly

implemented. Therefore, the proposed specifications ensure that the

consequences of a small break LOCA are not increased by allowing the

HPI pump discharge header to be cross-connected.

In addition, proposed Specification 3.4.7 requires new

operability requirements for the main steam atmospheric dump valves.

These operability requirements do not impact the probability or

consequences of any accident. The proposed specification for the

atmospheric dump valves provides additional assurance that these

valves will be operable in the event of a small break LOCA.

In summary, the proposed Technical Specifications provide

adequate controls to assure that operability of the HPI System is

maintained in a manner consistent with the requirements of the

design basis accidents. Therefore, it is concluded that this

amendment request will not significantly increase the probability or

consequences of an accident previously evaluated.

(2) Create the possibility of a new or different kind of

accident from any kind of accident previously evaluated:

No. Of the proposed substantive changes, only cross-connection

of the HPI pump discharge header represents any change to the way in

which the facility is normally operated. Operation with the

discharge header cross-connected is not a new configuration, as it

has always been used for HPI pump testing both at power and during

shutdown conditions. Potential failure modes have already been

considered as described earlier. No new initiating events or

potentially unanalyzed conditions have been created. Therefore, this

proposed amendment will not create the possibility of any new or

different kind of accident.

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

No. The HPI restrictions associated with the proposed Technical

Specifications are supported by analyses which demonstrate that the

acceptance criteria of 10 CFR 50.46 are not violated for any small

break LOCA. These analyses were performed in accordance with the

Evaluation Model described in FTI topical report BAW-10192P.

Therefore, it is concluded that the proposed amendment request will

not result in a significant decrease in the margin of safety.

Duke has concluded, based on the above, that there are no

significant hazards considerations involved in this amendment

request.

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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

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

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

Pennsylvania

Date of amendment request: March 10, 1997

Description of amendment request: The proposed amendment would

modify Technical Specification 3.4.5, ``Steam Generators,'' and

associated Bases to allow repair of steam generator tubes by

installation of sleeves with the tungsten inert gas (TIG) welded sleeve

developed by ABB Combustion Engineering. In addition, the proposed

amendment would delete the option for using the kinetic sleeving

methodology previously approved for use at Beaver Valley, but is not

currently recommended by Framatome Technologies, Inc.

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 amendment allows the ABB Combustion Engineering

(ABB/CE) tungsten inert gas (TIG) welded tubesheet sleeves and tube

support plate sleeves to be used as an alternate steam generator

tube repair method. The sleeve configuration was designed and

analyzed in accordance with the criteria of Regulatory Guide (RG)

1.121 and Section III of the ASME [American Society of Mechanical

Engineers] Code. Fatigue and stress analyses of the sleeved tube

assemblies produce acceptable results for both types of sleeves as

documented in ABB/CE Topical Report CEN-629-P, Revision 02 and CEN-

629-P Addendum 1. Mechanical testing has shown that the structural

strength of the sleeves under normal, faulted, and upset conditions

is within the acceptable limits specified in RG 1.121. Leakage rate

testing for the tube sleeves has demonstrated that primary to

secondary leakage is not expected during any plant condition. The

consequences of leakage through the sleeved region of the tube is

fully bounded by the existing steam generator tube rupture (SGTR)

analysis included in the Updated Final Safety Analysis Report

(UFSAR).

The sleeves are designed to allow inservice inspection of the

pressure retaining portions of the sleeve and parent tube. Inservice

inspection is performed on all sleeves following installation to

ensure that each sleeve has been properly installed and is

structurally sound. Periodic inspections are performed in subsequent

refueling outages to monitor sleeve degradation on a sample basis.

The eddy current technique used for inspection will be capable of

detecting both axial and circumferential flaws. Specific guidance

for steam generator sleeve inspection is provided in the current

technical specification surveillance requirements. Tubes that

contain defects in a sleeve, which exceed the repair limit, will be

removed from service. This ensures that sleeve and tube structural

integrity is maintained.

The proposed TS change to support the installation of TIG welded

sleeves does not adversely impact any previously evaluated design

basis accident. The effect of sleeve installation on the performance

of the SG [steam generator] was analyzed for heat transfer, flow

restriction, and steam generation capacity. The sleeves reduce the

risk of primary to secondary leakage in the SG. The installation of

ABB/CE sleeves results in a hydraulic flow restriction that is

dependent on the number and types of sleeves installed. The

reduction in primary system flow rate is a small percentage of the

flow rate reduction seen from plugging one tube and is a preferable

alternative when considering core margins based on minimum reactor

coolant system flow rates. The sleeving installation will result in

a resistance to primary coolant flow through the tube for other

evaluated accidents. The results of the analyses and testing, as

well as industry operating experience, demonstrate that the sleeve

assembly is an acceptable

[[Page 19830]]

means of maintaining tube integrity. In summary, installation of

sleeves does not substantially affect the primary system flow rate

or the heat transfer capability of the steam generators.

Installation of the sleeves can be used to repair degraded tubes

by returning the condition of the tubes to their original design

basis condition for tube integrity and leak tightness during all

plant conditions. The tube bundle overall structural and leakage

integrity will be increased with the installation of the sleeves

reducing the risk of primary to secondary leakage in the SG while

maintaining acceptable reactor coolant system flow rates. Therefore,

sleeving will not increase the probability of occurrence of an

accident previously evaluated.

Removal of the kinetically welded sleeve process as an approved

SG tube repair methodology will have no effect on plant operations.

There are currently no kinetically welded sleeves installed in the

steam generators. Had there been, plant operations would have still

been bounded by the existing SGTR analysis in the UFSAR.

Therefore, these proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The implementation of the proposed sleeving process will not

introduce significant or adverse changes to the plant design basis.

Stress and fatigue analyses of the repair has shown the ASME Code

Section III and RG 1.121 allowable values are met. Implementation of

TIG welded sleeving maintains overall tube bundle structural and

leakage integrity at a level consistent with that of the originally

supplied tubing. Leak and mechanical testing of the sleeves support

the conclusions that the sleeve retains both structural and leakage

integrity during all conditions. Repair of a tube with a sleeve does

not provide a mechanism that would result in an accident outside of

the area affected by the sleeve.

Any hypothetical accident as a result of potential tube or

sleeve degradation in the repaired portion of the tube is bounded by

the existing SGTR analysis. The SGTR analysis accounts for the

installation of sleeves and the impact on current plugging level

analyses. The sleeve design does not affect any other component or

location of the tube outside of the immediate area repaired.

The current primary to secondary leakage limit ensures that SG

tube integrity is maintained in the event of an MSLB [main steam

line break] or LOCA [loss-of-coolant accident]. The limit will

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

the occurrence of an unexpected single crack resulting in excessive

tube leakage. The leakage limit also provides for early detection

and a plant shutdown prior to a postulated crack reaching critical

crack lengths for MSLB conditions.

Inservice inspections are performed following sleeve

installation to ensure proper weld fusion has occurred to maintain

structural integrity. The post installation inspection also serves

as baseline data to be used for comparison during future

inspections. Periodic eddy current inspections monitor the pressure

retaining portions of the sleeve and parent tube for degradation.

Eddy current techniques will be employed that are sensitive to axial

and circumferential degradation.

Increasing the sample size of tubes repaired using either

sleeving process during each scheduled inservice inspection will

increase the monitoring of these tubes for any further degradation.

The improved monitoring and evaluation of the tube and the sleeves

assures tube structural integrity is maintained or the tube is

removed from service.

Corrosion testing of typical sleeve-tube configurations was

performed to evaluate local stresses, sleeve life, and resistance to

primary and secondary side corrosion. The tests were performed on

stress relieved and as-welded (non-stress relieved) sleeve-tube

joints. Using the corrosion test data in conjunction with finite

element analyses of the local stress, the stress relieved joint life

was determined to be in excess of 40 years. The ABB/CE TIG welded

sleeve operating experience in the industry has shown no sleeve

failures due to service induced degradation in sleeves that were

installed with acceptable inspection results. This experience

includes the stress relieved and as-welded sleeve configurations.

All sleeves will be stress relieved as specified in the topical

report.

Removal of the kinetically welded sleeve process as an approved

SG tube repair methodology and not completing the additional

corrosion testing necessary to establish the design life for the

kinetically welded sleeve in the presence of a crevice will not

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

any accident previously evaluated.

Repair of an SG tube with a kinetically welded sleeve would not

have provided a mechanism that resulted in an accident outside of

the area affected by the sleeve. Any hypothetical accident as a

result of potential tube or sleeve degradation in the repaired

portion of the tube would have been bounded by the existing SGTR

analysis. The SGTR analysis accounts for the installation of sleeves

and the impact on current plugging level analyses. The sleeve design

does not affect any other component or location of the tube outside

of the immediate area repaired. Furthermore, there are currently no

kinetically welded sleeves installed in either plant.

Therefore, the proposed changes do not create the possibility of

a new or different type of accident from any accident previously

evaluated.

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

of safety?

The TIG welded sleeving repair of degraded steam generator tubes

has been shown by analysis to restore the integrity of the tube

bundle to its original design basis condition. The safety factors

used in the design of the sleeves for the repair of degraded tubes

are consistent with the safety factors in the ASME Boiler and

Pressure Vessel Code Section III used in steam generator design. The

design of the ABB/CE SG sleeves has been verified by testing to

preclude leakage during normal and postulated accident conditions.

The portion of the installed sleeve assembly which represents

the reactor coolant pressure boundary can be monitored for the

initiation and progression of sleeve/tube wall degradation, thus

satisfying the requirement of RG 1.83. The portion of the SG tube

bridged by the sleeve joints is effectively removed from the

pressure boundary, and the sleeve then forms the new pressure

boundary. The sleeve enhances the safety of the plant by

reestablishing the protective boundaries of the steam generator.

Keeping the tube in service with the use of a sleeve instead of

plugging the tube and removing it from service increases the heat

transfer efficiency of the steam generator. During each scheduled

inservice inspection, each sleeve inspected and found to have

unacceptable degradation shall be removed from service.

The effect on the design transients and the accident analyses

have been revised based on the installation of sleeves equal to the

tube plugging level coincident with the minimum reactor coolant flow

rate. Evaluation of the installation of sleeves was based on the

determination that LOCA evaluations for the licensed minimum reactor

coolant flow bound the combined effect of tube plugging and sleeving

up to an equivalent of the actual plugging limit. Sleeving results

in a fractional amount of the plugging limitation of one tube and is

a preferable alternative when considering core margins based on

minimum reactor coolant system flow rates. The sleeving installation

will result in a resistance to primary coolant flow through the

tube. The primary coolant flow through the ruptured tube is reduced

by the influence of the installed sleeve; therefore, the

consequences to the public due to an SGTR event have not increased.

As SG sleeve removes an indication of a possible leak source

from the reactor coolant system (RCS) pressure boundary, eliminating

the potential of a primary-to-secondary leak. The structural

integrity of the tube is maintained by the sleeve and sleeve-to-tube

joint.

Installation of either tube sheet or tube support plate sleeves

will increase the protective boundaries of the steam generators and

will not reduce the margin of safety.

Removal of the kinetically welded sleeve process as an approved

SG tube repair methodology will not result in a reduction in the

margin of safety. There are currently no kinetically welded sleeves

installed in either plant. SG tube integrity will be maintained by

applying an alternate NRC approved repair methodology or removing

the SG tube from service by plugging.

Therefore, the proposed 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.

[[Page 19831]]

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

663 Franklin Avenue, Aliquippa, PA 15001

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

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

NRC Project Director: John F. Stolz

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

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

Pennsylvania

Date of amendment request: March 10, 1997

Description of amendment request: The proposed amendment would

revise Technical Specifications 3.4.5, ``Steam Generators,'' and

associated Bases to allow repair of steam generator tubes by

installation of sleeves with the Electrosleeving process developed by

Framatome Technologies, Inc. (FTI).

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 Electrosleeve configuration has been designed and analyzed

in accordance with the requirements of the ASME [American Society of

Mechanical Engineers] Code. The applied stresses and fatigue usage

for the Electrosleeve are bounded by the limits established in the

ASME Code. Minimum material property values are used for the

structural and plugging limit analysis. Mechanical testing has shown

that the structural strength of nickel Electrosleeves under normal,

upset, and faulted conditions provides margin to the acceptance

limits. These acceptance limits bound the most limiting (3 times

normal operating pressure differential) burst margin recommended by

Regulatory Guide 1.121. Leakage testing has shown that the

Electrosleeve is essentially leaktight during all plant conditions.

The Electrosleeve nominal wall thickness depth-based plugging

limit is determined using the guidance of Regulatory Guide 1.121 and

the pressure stress equation of Section III of the ASME Code. The

limiting requirement of Regulatory Guide 1.121 for the

Electrosleeve, which applies to part through wall degradation, is

the minimum acceptable wall thickness to maintain a safety factor of

three against tube failure under normal operating conditions. A

bounding set of design and transient loading input conditions was

used for the minimum wall thickness evaluation in the generic

evaluation. Evaluation of the minimum acceptable wall thickness for

normal, upset and postulated accident condition loading per the ASME

Code indicates these conditions are bounded by the design minimum

wall thickness.

Bounding tube wall degradation growth rate per cycle and

nondestructive examination uncertainty has been assumed for

determining the Electrosleeve technical specification plugging

limit. Electrosleeve wall degradation extent determined by

nondestructive examination, which would require plugging

Electrosleeved tubes, is developed using the guidance of Regulatory

Guide 1.121 and is defined in FTI Topical Report BAW-10219P,

Revision 1, to be 20% throughwall of the nominal sleeve wall

thickness.

The effect of Electrosleeving and plugging will remain below the

plugging limit assumed in the UFSAR [Updated Final Safety Analysis

Report]. The proposed change will not increase the consequences of

these accidents.

The results of the analyses and testing demonstrate that the

Electrosleeve is an acceptable means of maintaining tube integrity.

Further, per Regulatory Guide 1.83 recommendations, the

Electrosleeved tube can be monitored through periodic inspections

with present NDE [nondestructive examination] techniques. These

measures demonstrate that installation of Electrosleeves spanning

degraded areas of the tube will restore the tube to a condition

consistent with its original design basis.

Since the main steamline break post-accident primary-to-

secondary leakage is not increased by the presence of

Electrosleeves, the consequences of an accident previously evaluated

in the UFSAR are not increased. Conformance of the Electrosleeve

design with the applicable sections of the ASME Code and results of

the leakage and mechanical tests support the conclusion that

installation of Electrosleeves does not increase 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?

Electrosleeving will not adversely affect any plant component.

Stress and fatigue analysis of the repair has shown that the ASME

Code and Regulatory Guide 1.121 criteria are not exceeded.

Implementation of Electrosleeving maintains overall tube bundle

structural and leakage integrity at a level consistent with that of

the original tubing during all plant conditions. Leak and mechanical

testing of Electrosleeves support the conclusions of the

calculations that each Electrosleeve retains both structural and

leakage integrity during all conditions. Electrosleeving of tubes

does not provide a mechanism resulting in an accident outside of the

area affected by the Electrosleeves. Any accident resulting from

potential tube or Electrosleeve degradation in the repaired portion

of the tube is bounded by the existing tube rupture accident

analysis.

Implementation of Electrosleeving will reduce the potential for

primary-to-secondary leakage while not significantly impacting

available primary coolant flow area in the event of a LOCA. By

effectively isolating degraded areas of the tube through repair, the

potential for steamline break leakage is reduced. These degraded

intersections now are returned to a condition consistent with the

Design Basis. While the installation of an Electrosleeve reduces

primary coolant flow, the reduction is far below that caused by

plugging. Greater primary coolant flow area is maintained through

Electrosleeving versus plugging. Therefore, the possibility of a new

or different kind of accident from any accident previously evaluated

is not created.

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

of safety?

The Electrosleeve repair of degraded steam generator tubes has

been shown by analysis to restore the integrity of the tube bundle

consistent with its original design basis condition. The tube/

Electrosleeve operational and faulted condition stresses are bounded

by the ASME Code requirements and the Electrosleeved tubes are

leaktight. The safety factors used in the design of Electrosleeves

for the repair of degraded tubes are consistent with the safety

factors in the ASME Code used in steam generator design. The

portions of the installed Electrosleeve assembly which represent the

reactor coolant pressure boundary can be monitored for the

initiation and progression of Electrosleeve/tube wall degradation,

thus satisfying the requirements of Regulatory Guide 1.83. The

portion of the tube bridged by the Electrosleeve is effectively

removed from the pressure boundary, and the Electrosleeve then forms

the new pressure boundary. The areas of the Electrosleeved tube

assembly which require inspection are defined in Framatome

Technologies Inc. Topical Report BAW-10219P, Revision 1.

In addition, since the installed Electrosleeve represents a

portion of the pressure boundary, a baseline inspection of these

areas is required prior to operation with Electrosleeves installed.

The effect of sleeving on the design transients and accident

analyses has been reviewed based on the installation of

Electrosleeves up to the level of steam generator tube plugging

coincident with the minimum reactor coolant flow rate and UFSAR and

has been found acceptable.

It is concluded that the proposed license amendment request does

not result in a significant reduction in the margin of safety as

defined in the UFSAR or technical specifications.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001

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

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

NRC Project Director: John F. Stolz

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

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

Connecticut

Date of amendment request: March 27, 1997

[[Page 19832]]

Description of amendment request: The proposed changes to the

Technical Specifications (TSs) would modify the limiting condition for

operation (LCO) and surveillance requirements (SR) for the ultimate

heat sink. The ultimate heat sink for Millstone Unit No. 2 is the Long

Island Sound that transfers heat from safety-related systems during

normal and accident conditions. Specifically, TS LCO 3.7.11 would be

changed to indicate that the ultimate heat sink is operable at a water

temperature of less than or equal to 75 deg.F instead of an average

value. TS SRs 4.7.11.a and .b would also delete the use of average when

verifying the water temperature and delete the reference to a specific

monitoring location, the Unit No. 2 intake structure. These proposed

changes do not change the ultimate heat sink temperature limit, which

remains at a maximum of 75 deg.F.

The TS Bases 3/4.7.11 would also be modified to reflect the above

changes and to identify the various locations that the ultimate heat

sink temperature can be measured.

Basis for proposed no significant Hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes remove the reference to a monitoring

location where the temperature of the ultimate heat sink is measured

and eliminate the use of an average ultimate heat sink temperature.

The instruments used provide information to the operators which will

permit them to ensure that the plant is operated within the design

basis of the plant. The subject instruments will provide an accurate

representation of the ultimate heat sink temperature. This role is

passive; thus, these instruments cannot initiate or mitigate any

accident.

The locations used to monitor the ultimate heat sink temperature

will be maintained in the Bases. This is a licensee controlled

document which is maintained under the requirements of 10CFR50.59.

The details being removed from the Technical Specifications are not

assumed to be an initiator of any analyzed event. Since any changes

to the relocated details will be evaluated per 10CFR50.59, any

possible increase in the probability or consequences of an accident

previously evaluated will be addressed.

The proposed changes do not revise the ultimate heat sink

temperature limit of 75 deg.F. The current analysis is based on the

ultimate heat sink temperature limit of 75 deg.F. Therefore, there

is no effect on the consequences of any accident previously

evaluated.

Thus, the license amendment request does not impact the

probability of an accident previously evaluated nor does it involve

a significant increase in the consequences of an accident previously

evaluated.

2. Created the possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes remove the reference to a monitoring

location where the temperature of the ultimate heat sink is measured

and eliminate the use of an average ultimate heat sink temperature.

The instruments used provide information to the operators which will

permit them to ensure that the plant is operated within the design

basis of the plant. The subject instruments will provide an accurate

representation of the ultimate heat sink temperature. This role is

passive, thus, these instruments cannot initiate or mitigate any

accident.

The proposed changes will not alter the plant configuration (no

new or different type of equipment will be installed) or require any

new or unusual operator actions. They do not alter the way any

structure, system, or component functions and do not alter the

manner in which the plant is operated. The proposed changes do not

introduce any new failure modes. They will not alter assumptions

made in the safety analysis and licensing basis.

The locations used to monitor the ultimate heat sink temperature

will be maintained in the Bases. This is a licensee controlled

document which is maintained under the requirements of 10CFR50.59.

Thus, adequate control of information will be ensured.

Therefore, the changes will 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 remove the reference to a monitoring

location where the temperature of the ultimate heat sink is measured

and eliminate the use of an average ultimate heat sink temperature.

They do not change the ultimate heat sink temperature limit of 75

deg.F, which is assumed by the current analysis. Therefore, there is

no effect on the consequences of any accident previously evaluated

and no significant impact on offsite doses associated with

previously evaluated accidents. Thus, there is no significant

reduction in the margin of safety for the design basis accident

analysis. The license amendment request does not result in a

reduction of the margin of safety as defined in the Bases for

Technical Specification 3.7.11. The instruments used provide

information to the operators which will permit them to ensure that

the plant is operated within the design basis of the plant. The

subject instruments will provide an accurate representation of the

ultimate heat sink temperature. The proposed changes do not alter

the way any structure, system, or component functions and do not

alter the manner in which the plant is operated. They do not have

any impact on the protective boundaries (e.g., fuel matrix and

cladding, reactor coolant system pressure boundary, and primary and

secondary containment), or on the safety limits for these

boundaries.

The locations used to monitor the ultimate heat sink temperature

will be maintained in the Bases. The Bases are a licensee controlled

document which is maintained under the requirements of 10CFR50.59.

Since any future changes to this license controlled document will be

evaluated per the requirements of 10CFR50.59, any possible reduction

(significant or insignificant) in a margin of safety will be

addressed.

Thus, the license amendment request 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: Learning Resources Center,

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

Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385

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: March 31, 1997

Description of amendment request: The proposed amendment would

modify Technical Specification Surveillance Requirement 4.7.1.2.1.b

which requires the testing of the auxiliary feedwater motor-driven and

turbine-driven pumps on recirculation flow at least once per 92 days.

The proposed amendment would also makes changes to the appropriate

Bases section.

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 has reviewed the proposed changes in accordance with 10CFR

50.92 and has concluded that the changes do 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:

[[Page 19833]]

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed changes to Technical Specification Surveillance

4.7.1.2.1.b to increase the required test parameter for the motor

driven pumps from 1460 psid to 1468 psid and replacing the current

parameters for the motor driven and turbine driven pumps from

differential pressure measured in psid [pounds per square inch

differential] to total head measured in feet are consistent with

equipment design criteria and does not significantly increase the

probability of an accident previously evaluated.

The proposed changes to increase the required test parameter for

the motor driven pumps from 1460 psid to 1468 psid and replacing the

current parameters for the motor driven and turbine driven pumps

from differential pressure measured in psid to total head measured

in feet provides the necessary assurance that the pumps will

function as required in accident analyses and does not significantly

increase the consequence of an accident previously evaluated.

The moving of the reference to Specification 4.0.5 in order to

clarify that it applies to the testing of the motor driven and

turbine driven pumps and the modifications to the bases section are

administrative and do not involve a significant increase in the

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 Surveillance

4.7.1.2.1.b to increase the required test parameter for the motor

driven pumps from 1460 psid to 1468 psid and replacing the current

parameters for the motor driven and turbine driven pumps from

differential pressure measured in psid to total head measured in

feet does not change the operation of the auxiliary feedwater system

or any of its components during normal or accident evaluations.

The moving of the reference to Specification 4.0.5 in order to

clarify that it applies to the testing of the motor driven and

turbine driven pumps and the modifications to the bases section are

administrative and do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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 change to Technical Specification Surveillance

4.7.1.2.1.b to increase the referenced total head of the motor

driven auxiliary feedwater pumps during surveillance testing

provides an acceptable margin between the required surveillance and

design pump performance to provide assurance that the pumps will

operate consistent with system evaluations and does not involve a

significant reduction in a margin of safety.

The change in the referenced units from differential pressure

measured in psid to total head measured in feet for the motor driven

auxiliary and turbine driven auxiliary feedwater pumps during

surveillance testing is to account for the effect of water density

on pump performance during each test and does not involve a

significant reduction in a margin of safety.

The moving of the reference to Specification 4.0.5 in order to

clarify that it applies to the testing of the motor driven and

turbine driven pumps and the modifications to the bases section are

administrative and 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: March 31, 1997

Description of amendment request: The proposed amendment would

separate the required testing of motor-operated valve thermal overload

protection into two new 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 has reviewed the proposed change in accordance with

10CFR50.92 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 10CFR50.92(c) are not

satisfied. The proposed change does not involve a SHC because the

change would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed changes to the surveillance testing of the motor-

operated valve thermal overload protection are consistent with

equipment design criteria and performing surveillance testing does

not significantly increase the probability of an accident previously

evaluated. The proposed changes to the surveillance testing provides

the necessary assurance that the motor operated valve thermal

overload protection will function as required and does not involve a

significant increase in the 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 the surveillance testing of the motor-

operated valve thermal overload protection does not change the

operation of any system or system component during normal or

accident evaluations.

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 proposed changes to the surveillance testing of the motor-

operated valve thermal overload protection are administrative in

that the changes to the surveillance only clarify that following

maintenance on the motor starter, a channel calibration is required

only on that valve. The surveillance continues to require periodic

representative sample testing.

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 06141-0270NRC Deputy Director: Phillip F. McKee

[[Page 19834]]

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: March 4, 1997

Description of amendment request: The amendments would modify the

Emergency Core Cooling System (ECCS) surveillance test acceptance

criteria in Technical Specification 3/4.5.2 for the Centrifugal

Charging (CH) and the Safety Injection (SI) pumps. The changes to the

specified flow values would account for system alignments that effect

the suction pressure to the pumps. In the recirculation mode, increased

flow occurs when the CH and SI pumps take suction from the discharge of

the Residual Heat Removal pumps.

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 evaluations performed by Westinghouse determined

that, with the proposed changes, the subject pumps remain operable

and the safety analyses criteria remain valid.

Previous conclusions under LCR [License Change Request] 91-03

evaluating the consequences of the LOCA [loss-of-coolant-accident]

considered in the Salem Units 1 & 2 licensing basis remain

unchanged. With respect to the LOCA, the Peak Cladding Temperature

(PCT) continues to conform to the 10CFR50.46 guidelines of less than

2200*F. Evaluation of LOCA mass and energy releases previously found

acceptable remain valid. Decreasing the acceptance window to

accommodate the potential of an increase to pump runout flow,

assures that the current limits on pump runout flows continue to be

met. This change ensures pump integrity is maintained during the

accident. The reduction of the flow by throttling valves to

compensate for the potential suction boost remains within the

current analyses and therefore more conservative values are being

proposed. Additionally, the proposed change balances the pump flows

more appropriately by differentiating between the hot and cold leg

alignments. Flow to the reactor core is unaffected by the very

slight reduction in the upper flow limits. Since the design

limitations continue to be met and the integrity of the reactor

coolant system pressure boundary is not challenged, offsite dose

assumptions and calculations remain valid. Further, the ECCS is

post-accident mitigation system and probability of an accident is

not increased by this proposed change. Lastly, the correction of

double use of the word ``the'' in Salem Unit 1 Technical

Specification section 4.5.2.h.1.a is of editorial nature.

Based on the above information, the proposed changes do not

increase the risk 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. No new single failures are initiated. The proposed

changes will therefore not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed change addresses suction boost by changing the

Technical Specification surveillance acceptance criteria. The

typographical correction is of editorial nature.

3. The proposed change does not involve a significant reduction

in a margin of safety. The evaluation of LOCA accident analysis

previously performed by Westinghouse continues to be met and

verifies that, with the proposed changes to the TS, plant operations

will be maintained within the bounds of safe, analyzed conditions as

defined in the UFSAR [Updated Final Safety Analysis Report] and that

conclusions presented in the UFSAR remain valid. The peak cladding

temperatures (PTC) remains unchanged as no effective differences in

the operating parameters have occurred. The typographical correction

is of editorial nature. The proposed changes will therefore not

reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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

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: March 7, 1997

Description of amendments request: The proposed amendments would

allow operability testing for the containment isolation valves listed

in Table 3.6-1 of the Technical Specifications during a defueled

status. 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. Involve a significant increase in the probability or

consequences of an accident previously evaluated.]

The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated in

the FSAR [Final Safety Analysis Report]. The proposed changes have

no impact on the probability of an accident. The containment

isolation valves will continue to require operability testing.

Allowing the testing to be performed when the unit is in a defueled

status will have no impact on any accidents previously evaluated.

The net effect of these changes is not significant and, as a result,

does not involve a significant increase in the consequences 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 the Technical Specifications do not

increase the possibility of a new or different kind of accident than

any accident already evaluated in the FSAR. No new limiting single

failure or accident scenario has been created or identified due to

the proposed changes. Safety-related systems will continue to

perform as designed. The proposed changes do not create the

possibility of a new or different kind of accident from any

previously evaluated.

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

The proposed changes do not involve a significant reduction in

the margin of safety. Although, as a result of these proposed

changes, the containment isolation valves could be tested for

operability while the unit is in a defueled state, there is no

impact in the accident analyses. These proposed changes are

technically consistent with the requirements of NUREG-1431, Revision

1 which has already received the requisite review and approval of

the NRC staff. Thus the proposed 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 location: Houston-Love Memorial Library,

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

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

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

Alabama 35201

NRC Project Director: Herbert N. Berkow

[[Page 19835]]

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

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

Date of amendment request: August 22, 1996, as supplemented on

March 28, 1997 (TS 96-02)

Description of amendment request: The proposed changes would revise

Section 3.6.5 of the Sequoyah Technical Specifications (TS) and

associated Bases to lower the minimum TS ice basket weight of 1,155

pounds to 1,071 pounds. This would reduce the overall weight of ice

required in the ice condenser from 2,245,320 pounds to 2,082,024

pounds. The TVA license amendment request also proposed to extend the

chemical analysis surveillance interval for the ice condenser ice bed

from 12 months to 18 months based on the provisions of Generic Letter

93-05.

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.

TVA proposes to modify the SQN Unit 1 and Unit 2 TSs [Technical

Specifications] to revise Surveillance Requirement (SR) 4.6.5.1.d to

lower SQN's minimum TS basket weight from 1,155 pounds (lbs) to

1,071 lbs, thus lowering the overall ice condenser weight from

2,245,320 lbs to 2,082,024 lbs.

The ice condenser system is provided to absorb thermal energy

release following a loss-of-coolant accident (LOCA) or high energy

line break (HELB) and to limit the peak pressure inside containment.

The current containment analysis for SQN is based on a minimum of

993 lbs of ice per basket evenly distributed throughout the ice

condenser at the end of an 18-month refueling cycle. The revised

containment analysis shows that for the predicted sublimation rate

of 15 percent for 18 months, an average basket weight of 922 lbs at

the end of the 18-month period would ensure containment design

pressure is not exceeded.

Based on TVA's evaluation and the revised containment analysis,

TVA considers the reduction of ice weight to be acceptable for

satisfying the safety function of the ice condenser for an 18-month

ice weighing interval. Therefore, the proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

TVA is also proposing to extend the surveillance interval as it

pertains to the ice bed chemical analysis. Based on test results,

both at SQN and the industry, the average boron concentration and pH

changes are minimal; therefore, this change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Elimination of the temperature at which the pH of the ice bed is

determined is an administrative change. Future testing will be

accomplished in accordance with American Society for Testing and

Materials Standards recommendations. Therefore, this change cannot

increase the probability of an accident and the consequences of an

accident will not increase.

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

from any previously analyzed.

TVA's request to lower the TS limit for ice weight at the start

of the surveillance interval will not result in a new or different

kind of accident from that previously analyzed in SQN's Final Safety

Analysis

Report. SQN's ice condenser serves to limit the peak pressure

inside containment following a LOCA. TVA has evaluated the revised

containment pressure analysis for SQN (Enclosure 4, Westinghouse

WCAP-12455, Revision 1) and determined that sufficient ice would be

present at all times to keep the peak containment pressure below

SQN's containment design pressure of 12 pounds per square inch gage

(psig). Therefore, this change would not result in a new or

different kind of accident from any previously analyzed.

The proposed reduced testing frequency of the chemical

composition of the ice bed does not change the manner in which the

plant is operated. Additionally, the ice condenser is a passive

system that reacts to an accident, but does not support plant

operation on a daily basis. The reduced testing frequency of the ice

bed chemical composition does not generate any new accident

precursors; therefore, the possibility of a new or different kind of

accident from any previously analyzed is not created.

Elimination of the temperature at which the pH of the ice bed is

determined is an administrative change. This change cannot create

the possibility of a new or different kind of accident.

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

The ice condenser system is provided to absorb thermal energy

release following a LOCA and to limit the peak pressure inside

containment. The current ice condenser analysis for SQN is based on

a minimum of 993 lbs of ice per basket. The revised containment

analysis changes the minimum ice weight assumed in the analysis to

922 lbs per basket.

The revised containment analysis shows that using an average

basket weight of 1,071 lbs and a sublimation allowance of 15

percent, all bays would have an average basket weight of 922 lbs at

the end of the 18-month surveillance interval. The revised analysis

utilizes new mass and energy releases (refer to Westinghouse WCAP-

10325-P-A), which substantially delays ice-bed meltout and limits

the initial containment peak pressure to approximately 7.15 psig

during the blowdown phase. The ice-bed meltout delay allows the

second containment pressure peak, which is driven mainly by the

decay heat, to be limited to approximately 11.45 psig, which is

below the containment design pressure of 12 psig.

Based on TVA's evaluation and the revised containment analysis,

TVA considers the reduction of the average basket weight to be

acceptable for satisfying the safety function of the ice condenser

for the current 18-month interval. Therefore, the proposed change

does not involve a significant reduction in the margin of safety.

The proposal to extend the surveillance from 12 to 18 months

does not change the boron concentration or pH requirements.

Experience at Duke Power Company, as stated in NUREG-1366, indicates

that these parameters do not change appreciably when verified every

9 months. SQN has a similar experience with a 12-month interval.

Since the boron concentration and the post-LOCA pH requirements

remain essentially the same, there is no reduction in the margin of

safety.

Elimination of the temperature at which the pH of the ice bed is

determined is an administrative change. Future testing will be

accomplished in accordance with ASTM recommendations. The difference

between the pH values determined at the current TS specified

temperature and the temperature currently recommended by the ASTM

standards is insignificant. Therefore, there is no reduction in the

margin of safety.

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

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: September 12, 1996

Description of amendment request: The proposed change to the

Technical Specifications is administrative in nature in that it would

add the NRC standard fire protection license condition to each unit's

Operating License and relocate the fire protection requirements from

the Technical Specifications to the Updated Final Safety Analysis

Report (UFSAR).

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:

[[Page 19836]]

Specifically, operation of Surry Power Station with the proposed

amendment will not:

1. Involve a significant increase in either the probability of

occurrence or consequences of any accident or equipment malfunction

scenario that is important to safety and which has been previously

evaluated in the UFSAR. The requirements of the Fire Protection

Program have not been changed by theproposed amendment. Relocation

of the Fire Protection Program requirements into the UFSAR and

station procedures does not decrease any portion of the program. The

same fire protection requirements exist as before the change.

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

than those previously evaluated in the safety analysis report. The

requirements of the Fire Protection Program have not been changed by

the proposed amendment. This is an administrative change to relocate

the Fire Protection Program requirements from the Technical

Specifications to the UFSAR and station procedures. Consequently,

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

accident previously evaluated has not been created.

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

Implementation of the Fire Protection Program requirements is

assured by the UFSAR and station procedures. Since the rogram is

being retained intact, there is no reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219

NRC Project Director: Mark Reinhart, Acting

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: January 16, 1997

Description of amendment request: The proposed amendments (Point

Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request

(TSCR) 191) would revise the minimum boron concentration required in

the refueling water storage tank(s)(RWST), boric acid storage tanks

(BAST), and safety injection (SI) accumulators during normal operation;

the minimum boron concentration of primary coolant during refueling

conditions; and the minimum boron concentration in the reactor when

positive reactivity could be added and/or boron dilution could occur

and containment integrity is not intact. These changes are necessary to

accommodate the planned extension of the operating cycle from 12 months

to 18 months. The licensee proposes to change TS 15.3.2, ``Chemical and

Volume Control System,'' TS 15.3.3, ``Safety Injection and Residual

Heat Removal Systems,'' TS 15.3.6, ``Containment System,'' TS 15.3.8,

``Refueling,'' and associated Bases.

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 this facility under the proposed Technical

Specifications will not create a significant increase in the

probability or consequences of an accident previously evaluated.

The probabilities of accidents previously evaluated are based on

the probability of initiating events for these accidents. Initiating

events for accidents previously evaluated are described in the PBNP

FSAR [final safety analysis report].

In effect, the proposed changes will result in: (1) higher boron

concentrations of primary coolant during refueling, and (2) higher

boron inventories in the RWSTs, BASTs, and SI accumulators. These

changes do not require hardware changes or changes to the operation

of accident-mitigating equipment. These changes relate to the

performance capability of particular accident mitigation systems;

equipment that is not postulated to cause accidents. Therefore,

these proposed changes do not cause an increase in the probabilities

of any accidents previously evaluated.

The consequences of accidents previously evaluated in the PBNP

FSAR are determined by the results of analyses that are based on

initial conditions of the plant, the type of accident, transient

response of the plant, and the operation and failure of equipment

and systems.

In effect, the proposed changes will result in: (1) higher boron

concentrations of primary coolant during refueling, and (2) higher

boron inventories in the RWSTs, BASTs, and SI accumulators. These

increased boron concentrations do not increase the probability that

engineered safety features equipment will fail, nor do these changes

affect the capability of this equipment to operate as required for

the accidents previously evaluated in the PBNP FSAR. These changes

do not require hardware changes or changes to the operation of

accident-mitigating equipment.

The consequential effects of a lower containment spray pH will

not affect the capability of the containment spray to remove

elemental iodine during design basis LOCA [loss-of-coolant accident]

accidents. Also, the consequential reduction in containment sump

water pH will not affect the fluid's capability to retain elemental

iodine, nor will it adversely increase the potential corrosion rates

for materials inside containment if the sump water is sprayed into

containment during the recirculation phase of a LOCA.

Another consequence of injecting a higher concentration boric

acid solution into the core during a LOCA may be an abbreviated

onset to boron precipitation in the post-LOCA core. An incremental

change in the boron injection concentration would not have

significant effect on the postulated onset, but each core reload

safety evaluation will continue to verify that the existing

emergency operating procedures accommodate the potential for boron

precipitation.

Therefore, this proposed license amendment does not affect the

consequences of any accident previously evaluated in the PBNP FSAR,

because the factors that are used to determine the consequences of

accidents are not changed.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of a new or

different kind of accident from any previously evaluated.

New or different kinds of accidents can only be created by new

or different accident initiators or sequences. New and different

types of accidents (different from those that were originally

analyzed for Point Beach) have been evaluated and incorporated into

the licensing basis for PBNP. Examples of different accidents that

have been incorporated into the PBNP licensing basis include

anticipated transients without scram and station blackout.

The changes proposed by this TSCR do not create any new or

different accident initiators or sequences because these changes to

minimum boron concentrations will not cause failures of equipment or

accident sequences different than the accidents previously analyzed.

No new equipment interfaces are created, and no new materials or

fluids are introduced. The incremental increase in boron

concentrations will not create a failure mechanism not previously

known and evaluated. Therefore, these proposed TS changes do not

create the possibility of an accident of a different type than any

previously evaluated in the PBNP FSAR.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are based on the design

and operation of the reactor and containment and the safety systems

that provide their protection. Plant safety margins are established

through Limiting Conditions for Operation, Limiting Safety System

Settings and Safety Limits specified in the Technical

Specifications. The proposed Technical Specification changes to

refueling water storage tank (RWST), SI accumulator, and BAST boron

inventory requirements have all been evaluated to preserve the

shutdown capability described in the associated bases (boration from

just critical, hot zero or full power, peak xenon with control rods

at the

[[Page 19837]]

insertion limit, to xenon-free cold shutdown with the highest worth

control rod assembly fully withdrawn). Similarly, the proposed TS

change to the minimum boron concentration of the primary coolant

system for refueling operations have been evaluated to preserve the

subcriticality margin described in the associated TS bases (i.e., 5%

[delta] k/k in the cold condition with all rods inserted).

Because there are no changes to any of these margins, the

proposed license amendment does not involve a reduction in any

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: January 21, 1997

Description of amendment request: The proposed amendments (Point

Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request

195) would revise TS Section 15.6.11, ``Radiation Protection Program,''

to update all references to 10 CFR Part 20, ``Standards for Protection

Against Radiation,'' to restore consistency between 10 CFR Part 20

regulations and the PBNP TS.

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 create a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed amendments are administrative in nature, providing

consistency between the Point Beach licenses and Commission

regulations. The amendments do not affect the operation or

maintenance of any PBNP structure[,] system or component. In

addition, the regulations and proposed changes provide more

conservative determinations of high radiation areas, thereby

potentially resulting in lower personnel radiation exposures during

normal operation and post accident. The consequences of an accident

related to personnel radiation exposures may be reduced.

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 are administrative only and do not

affect the operation or maintenance of any structure[,] system or

component at Point Beach Nuclear Plant. No new systems or components

are introduced. Therefore, no new accident initiators or sequences

result from any previously evaluated.

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

the proposed amendments will not create a significant reduction in a

margin of safety.

The proposed amendments are administrative and reflect

regulatory requirements that are more conservative than those

presently reflected in the PBNP Technical Specifications. These more

conservative requirements result in more conservative designation of

high radiation areas thereby providing additional margins of safety

related to the control of radiation exposures to personnel. No

structure[,] system or component at PBNP at PBNP is changed[,]

thereby maintaining the margins of safety for the operation of the

Point Beach Nuclear Plant.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: January 24, 1997

Description of amendment request: The proposed amendments (Point

Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request

(TSCR) 193) would revise TS 15.5.4, ``Fuel Storage,'' to increase fuel

assembly enrichment limits to 5.0 w/o U-235 while maintaining Keff in

the storage pools (spent fuel pool and new fuel storage racks) less

than 0.95.

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 this facility under the proposed Technical

Specifications will not create a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve a change to structures,

systems, or components which would affect the probability or

consequences of an accident previously evaluated in the PBNP Final

Safety Analysis Report (FSAR). The only relevant concern with

respect to increasing enrichment limits in the spent fuel pool and

new fuel storage racks is one of criticality. The proposed changes

use the same criticality limit used in the current Technical

Specifications. Therefore, margin to safe operation of Units 1 and 2

is maintained. The probability and consequences of an accident

previously evaluated are dependent on this criticality limit.

Because the limit will not change, the probability and consequences

of those accidents previously evaluated will not change.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes do not involve a change to plant design.

The proposed increase in spent fuel pool and new fuel storage racks

fuel assembly enrichment limits maintains the margin to safe

operation of Units 1 and 2 because the criticality limit for the

spent fuel pool and new fuel storage racks will not change. These

changes do not affect any of the parameters or conditions that

contribute to the initiation of any accidents. Because the

criticality limit remains the same, these changes have no effect on

plant operation, design, or initiation of any accidents. Therefore,

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

different kind of accident from any accident previously evaluated.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety.

The proposed changes maintain the margin to safe operation of

Units 1 and 2. The margin of safety is based on the criticality

limit of the spent fuel pool and the new fuel storage racks. Because

this limit will not change, the margin of safety will not be

affected. Therefore, the proposed changes will not create 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

[[Page 19838]]

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: February 12, 1997, as supplemented on

March 11, 1997

Description of amendment request: The proposed amendments (Point

Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request

196) would relocate turbine overspeed protection specifications,

limiting conditions for operation, surveillance requirements, and

associated bases from TS Section 15.3.4, ``Steam and Power Conversion

System,'' and Section 15.4.1, ``Operational Safety Review,'' to the

Final Safety Analysis Report (FSAR) in accordance with Generic Letter

95-10, ``Relocation of Selected Technical Specifications Requirements

Related to Instrumentation.''

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 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 amendments administratively relocate turbine

overspeed protection Specifications to the Point Beach Final Safety

Analysis Report (FSAR). The Specifications will be transferred

verbatim, except for the turbine load limit with the crossover steam

dump system inoperable, which has already been evaluated under 10

CFR 50.59 and will be conservatively reduced. In addition, the

regulatory requirements of 10 CFR 50.55a, ``Codes and Standards, ''

will still apply to the relocated Specifications. Therefore,

operation of Point Beach Nuclear Plant in accordance with the

proposed amendments cannot create an increase in the probability or

consequences of an accident previously evaluated.

2. Operation of 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 administratively relocate Specifications

to the FSAR and in one case result in a more conservative operating

limit. Therefore, operation of Point Beach Nuclear Plant in

accordance with the proposed amendments cannot create a new or

different kind of accident from any accident previously evaluated.

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

proposed amendments will not create a significant reduction in a

margin of safety.

The proposed changes are administrative in nature. There is no

physical change to the facility, its systems, or its operation,

except for the conservative reduction of the turbine load limit with

the crossover steam dump system inoperable which has already been

justified via 10 CFR 50.59. Therefore, operation of PBNP in

accordance with the proposed amendments cannot 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: 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: February 17, 1997; supersedes March 24,

1995, as supplemented by letter dated August 16, 1995, amendment

request.

Description of amendment request: This amendment request proposes

to revise Technical Specification 1.7, ``Containment Integrity,''

Technical Specification 3/4.6.1, ``Containment Integrity,'' and

Technical Specification 3/4.6.3, ``Containment Isolation Valves.''

These proposed changes would relocate Technical Specification Table

3.6-1, ``Containment Isolation Valves,'' to the Wolf Creek Generating

Station (WCGS) procedures. This proposed change is in accordance with

the guidance provided in Generic Letter 91-08, ``Removal of Component

Lists from Technical Specifications,'' dated May 6, 1991. In addition,

this request proposes that the August 16, 1996, supplemental submittal

that provided an additional footnote allowing an increased outage time

for certain component cooling water system valves be withdrawn. The

determination that the additional footnote is not required supersedes

the staff's proposed no significant hazards consideration determination

evaluation for the requested changes that was published on September

27, 1995 (60 FR 49949).

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 simplify the technical specifications, meet

the regulatory requirements for control of containment isolation,

and are consistent with the guidelines of GL 91-08. The procedural

details of Technical Specification Table 3.6-1 have not been

changed, but only relocated to a different controlling document. The

proposed changes are administrative in nature, should result in

improved administrative practices, and do not affect plant

operations.

The probability of occurrence of a previously evaluated accident

is not increased because this change does not introduce any new

potential accident initiating conditions. The consequences of an

accident previously evaluated is not increased because the ability

of containment to restrict the release of any fission product

radioactivity to the environment will not be degraded by this

change.

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 are administrative in nature, do not result

in physical alterations or changes to the operation of the plant,

and cause no change in the method by which any safety-related system

performs its function. Therefore, this proposed change will 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 administrative change to relocate Technical Specification

Table 3.6-1 to appropriate plant procedures does not alter the basic

regulatory requirements for containment isolation and will not

adversely affect containment isolation capability for Coordinator

credible accident scenarios. Adequate control of the content of the

table is assured by existing plant procedures.

The proposed relocation of Technical Specification Table 3.6-1

does not alter current technical specification requirements for

containment isolation valve operability. The LCO and Surveillance

Requirements would be retained in the revised technical

specifications. Therefore, the proposed change will not affect the

meaning, application, and function of the current technical

specification requirements for the valves in Table 3.6-1.

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

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

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

amendment request involves no significant hazards consideration.

[[Page 19839]]

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

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: March 18, 1997

Description of amendment request: This license amendment request

revises Technical Specification Surveillance Requirement 4.5.2.c to

clarify when a containment entry visual inspection is required. This

proposed change to reduce the visual inspection requirement to at least

once daily is in accordance with the guidance provided in Generic

Letter 93-05, ``Line-Item Technical Specifications Improvements to

Reduce Surveillance Requirements for Testing During Power 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.

Implementing the proposed change could potentially increase the

chances of loose debris (trash, rags, clothing, etc.) being left in

containment for some period of time greater than would be allowed

under current surveillance requirements. However, the proposed

change also clarifies that the visual inspection must be performed

at least once daily. Therefore, the period of time that debris could

be left uncontrolled inside containment would still be less than 24

hours. Based on work controls placed on material entry/exit into

containment and personnel training on housekeeping controls inside

containment, and the results of past surveillances, it is unlikely

that a significant amount of debris would be left uncontrolled

inside containment for this period of time. Also, based on

containment sump design, relatively small amounts of debris would

not be sufficient to cause a significant amount of blockage of the

sump screens.

The probability of occurrence of a previously evaluated accident

is not increased because the reduced frequency of the visual

inspection does not cause a significant impact on the possibility of

containment sump screen blockage. Therefore containment sump

operability is not affected by the proposed change. In addition, the

proposed change will not result in any changes to the design or

operation of any plant systems or components.

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

or different kind of accident from any accident previously

evaluated.

The proposed change decreases the frequency of performing a

visual inspection for loose debris in containment, but does not

result in a change to the design or operation of any plant system or

component. The purpose of the inspection is to ensure that there is

no loose debris, left in containment following a containment entry,

that could potentially block the containment sump screens during

LOCA conditions. Delaying this inspection until the last containment

entry each day will not result in a significant amount of debris

being left in containment, based on housekeeping practices

controlling the entry/removal of trash and material into/from

containment; training of employees to increase awareness of material

control in radiologically-controlled areas; and retaining the

requirement to perform a visual inspection at least once per day

when containment entries are made (during periods when containment

integrity is established), thereby ensuring that trash and debris

can be identified and removed on a daily basis (on days containment

entries are made).

Based on the above, this proposed change will 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 purpose of performing a visual inspection of areas affected

by a containment entry is to ensure any debris or trash generated by

the activity during the containment entry is identified and removed

from containment. This ensures that no trash or debris is left in

containment that could be transported to and block the containment

sump screens during LOCA conditions. Based on current material

control and housekeeping practices imposed on containment entry/

exit, and past inspection results, reducing the surveillance

requirement to a once per day basis, on days containment entries are

made, would not result in a significant amount of trash or debris

being left in containment following completion of the entry, and any

such material would be identified and removed prior to the end of

the day. The amount of trash or debris that could be left in

containment for a 24 hour period would be significantly less than

the amount that would be required to cause sump screen blockage

sufficient to affect sump performance. Therefore, the proposed

change will not result in a significant reduction in the margin of

safety of any plant system or equipment.

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

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: March 18, 1997

Description of amendment request: This license amendment request

revises Technical Specification Section 5.3.1, Fuel Assemblies, to

allow the use of an alternate zirconium based fuel cladding material,

ZIRLO. Wolf Creek Nuclear Operating Corporation (WCNOC) is planning to

insert Westinghouse fuel assemblies containing ZIRLO fuel rod cladding

during the ninth refueling outage, which is currently scheduled to

begin in late September 1997.

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 methodologies used in the accident analysis remain

unchanged. The proposed changes do not change or alter the design

assumptions for the systems or components used to mitigate the

consequences of an accident. Use of ZIRLO fuel cladding does not

adversely affect fuel performance or impact nuclear design

methodology. Therefore accident analyses are not impacted.

The operating limits will not be changed and the analysis

methods to demonstrate operation within the limits will remain in

accordance with NRC approved methodologies. Other than the changes

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

associated with this technical specification change. A safety

analysis will continue to be performed for each cycle to demonstrate

compliance with all fuel safety design basis.

VANTAGE 5H with IFMs fuel assemblies with ZIRLO clad fuel rods

meet the same fuel assembly and fuel rod design bases as other

VANTAGE 5H with IFMs fuel assemblies. In addition, the 10 CFR 50.46

criteria are applied to the ZIRLO clad rods. The use of these fuel

assemblies will not result in a change to the reload design and

safety analysis limits. The clad material is similar

[[Page 19840]]

in chemical composition and has similar physical and mechanical

properties as Zircaloy-4. Thus, the 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 use

of an assembly containing a combination of Zircaloy-4 and ZIRLO clad

fuel rods. Since the dose predictions in the safety analyses are not

sensitive to fuel rod cladding material, the radiological

consequences of accidents previously evaluated in the safety

analysis remain valid.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident or

malfunction of equipment important to safety previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

VANTAGE 5H with IFMs fuel assemblies with ZIRLO clad fuel rods

satisfy the same design bases as those used for other VANTAGE 5H

with IFMs fuel assemblies. All design and performance criteria

continue to be met and no new failure mechanisms have been

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

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

malfunction of equipment important to safety. The ZIRLO cladding

material offers improved corrosion resistance and structural

integrity.

The proposed changes do not affect the design or operation of

any system or component in the plant. The safety functions of the

related structures, systems or components are not changed in any

manner, 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 change to the facility or operating procedures,

and the safety functions and reliability of structures, systems and

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

possibility of a new or different kind of accident or malfunction of

equipment important to safety from any accident or malfunction of

equipment important to safety previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Use of ZIRLO cladding material does not change the VANTAGE 5H

with IFMs reload design and safety limits. The use of these fuel

assemblies will take into consideration the normal core operating

conditions allowed in the Technical Specifications. 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.

The use of Zircaloy-4, ZIRLO or stainless steel filler rods in

fuel assemblies will not involve a significant reduction in the

margin of safety because analyses using NRC approved methodologies

will be performed for each configuration to demonstrate continued

operation within the limits that assure acceptable plant response to

accidents and transients. These analyses will be performed using NRC

approved methods that have been approved for application to the fuel

configuration.

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

involving no significant hazards consideration.

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

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

of the original notice.

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

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

County, North Carolina

Date of amendments request: March 27, 1997

Description of amendments request: The proposed amendments would

revise the Technical Specifications for the Brunswick Steam Electric

Plant Units 1 and 2 to eliminate certain instrumentation response time

testing requirements in accordance with NRC-approved BWR Owners Group

Topical Report NEDO-32291-A, ``System Analysis for the Elimination of

Selected Response Time Testing Requirements.''Date of publication of

individual notice in Federal Register: April 1, 1997 (62 FR 15542)

Expiration date of individual notice: May 1, 1997

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph

M. Farley Nuclear Plant, Unit No. 1, Houston County, Alabama

Date of amendment request: March 25, 1997

Description of amendment request: The proposed amendment would

modify Technical Specification 3/4.4.9, ``Specific Activity,'' and

associated Bases to reduce the limit associated with dose equivalent

iodine-131. The steady-state dose equivalent iodine-131 limit would be

reduced by 40 percent from .5 [micro]Curie/gram to .3 [micro]Curie/

gram.

Date of publication of individual notice in Federal Register: April

4, 1997 (62 FR 16201)

Expiration date of individual notice: May 5, 1997

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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

[[Page 19841]]

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.

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 19, 1997, as

supplemented April 3, 1997.

Brief description of amendments: The amendments would delete the

24/48 Volt direct current (Vdc), batteries, battery chargers and

distribution systems from the Technical Specifications (TSs) for Unit

3, by adding a footnote to indicate that these TSs are only applicable

to Unit 2. All safety-related loads associated with the 24/48 Vdc

batteries for Unit 3 will be relocated to other safety-related battery

systems which are in the TSs.

Date of issuance: April 10, 1997

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

Amendment Nos.: 156 and 151

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

revised the Technical Specifications.

Date of initial notice in Federal Register: March 5, 1997 (62 FR

10088). The April 3, 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

April 10, 1997. No significant hazards consideration comments received:

No

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450

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

PointNuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: August 14, 1996, as supplemented

September 13, 1996

Brief description of amendment: The amendment revises Technical

Specification Sections 3.3 and 6.9.1.9; and the basis of Section 3.3,

3.6 and 3.10. The changes incorporate the best estimate approach into

the licensing basis for the Indian Point Unit No. 2 loss-of-coolant

accident analysis.

Date of issuance: March 31, 1997

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

30 days.

Amendment No.: 188

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

Technical Specifications.

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

4344) The September 13, 1996, supplemental letter did not change the

initial proposed no significant hazards consideration.The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated March 31, 1997.No significant hazards consideration comments

received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610

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

PointNuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: February 14, 1997, as

supplemented March 12, 1997.

Brief description of amendment: The amendment revises Technical

Specification Section 4.13-2 to allow a one-time extension of the

interval for steam generator tube inspection. Specifically, the date

for commencement of the steam generator tube inspection is extended

from April 14, 1997 to May 2, 1997.

Date of issuance: April 9, 1997

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

April 14, 1997.

Amendment No.: 189

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

Technical Specifications.

Date of initial notice in Federal Register: March 4, 1997 (62 FR

9816) The March 12, 1997, supplemental letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated April 9, 1997.No

significant hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610

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

Charlevoix County, Michigan

Date of application for amendment: November 7, 1996

Brief description of amendment: The amendment revised Technical

Specification 4.2.9, Service and Instrument Air System, to add an

additional air compressor.

Date of issuance: April 2, 1997

Effective date: Effective the date of issuance.

Amendment No.: 118

Facility Operating License No. DPR-6: Amendment revised the

Technical Specifications.

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

FR 66706) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 2, 1997.No significant

hazards consideration comments received: No.

Local Public Document Room location: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770

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: January 3, 1997, as

supplemented by letter dated March 20, 1997

Brief description of amendments: The amendments revise Technical

Specification Tables 3.3-2, 3.3-4, 3.3-5, 4.3-2 and Bases Sections 3/

4.3.1 and 3/4.3.2 to eliminate the safety injection signal on low steam

line pressure.

Date of issuance: April 3, 1997

Effective date: For Unit 1, as of the date of issuance to be

implemented before startup from the next refueling outage; For Unit 2,

as of the date of issuance to be implemented before startup from the

current refueling outage

Amendment Nos.: 158 and 150

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

revised the Technical Specifications.

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

4345) The March 20, 1997, letter provided clarifying information that

did not change the scope of the original January 3, 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 3, 1997.No significant hazards

consideration comments received: No

[[Page 19842]]

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of application for amendments: February 5, 1997

Brief description of amendments: The amendments reflect replacement

of the existing source and intermediate range nuclear instrumentation

with a new source range and wide range nuclear instrumentation system

that provides more channels and continuous coverage from the Source

Range to above the Power Range.

Date of issuance: March 31, 1997

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

30 days.

Amendment Nos.: 223, 223, 220

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

amendments revised the Technical Specifications.

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

FR 8796) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 31, 1997.No significant

hazards consideration comments received:

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: October 16, 1996

Brief description of amendment: The amendment changes the Appendix

A Technical Specifications by revising Table 4.3-1 to expand the

applicability for Core Protection Calculator (CPC) operability and to

allow the use of a cycle independent shape annealing matrix in the

CPCs.

Date of issuance: April 11, 1997

Effective date: April 11, 1997, to be implemented within 60 days

Amendment No.: 125

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

Technical Specifications.

Date of initial notice in Federal Register: February 12, 1997 (62

FR 6575) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 11, 1997No significant

hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 2, 1996 as supplemented by

letter dated February 4 and March 14, 1997

Brief description of amendment: The amendment changes the Technical

Specifications to reflect the approval for the licensee to use of the

new Containment Leakage Rate Testing Program as required by 10 CFR Part

50 Appendix J, Option B for Waterford Steam Electric Station, Unit 3.

Date of issuance: April 10, 1997

Effective date: April 10, 1997

Amendment No.: 124

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

Technical Specifications.

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

2189) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 10, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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

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

Date of application for amendment: December 9, 1996

Brief description of amendment: This amendment modifies technical

specifications for selected cycle-specific reactor physics parameters

to refer to the St. Lucie Unit 1 Core Operating Limits Report for

limiting values.

Date of issuance: April 1, 1997

Effective date: April 1, 1997

Amendment No.: 150

Facility Operating License No. NPF-16: Amendment revised the

Technical Specifications.

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

2189) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 1, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room location: Indian River Junior College

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

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

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

Connecticut

Date of applications for amendment: June 20, 1995, as supplemented

August 30, 1995, and January 17, 1996

Brief description of amendment: The amendment relocates the

applicable requirements of Technical Specification (TS) 3.6.3 for the

main steam line isolation valves (MSIVs) to TS 3.7.1.5, ``Main Steam

Line Isolation Valves.'' In addition, the Applicability section of TS

3.7.1.5 is revised to indicate that Specification 3.7.1.5 is applicable

in Mode 1 and in Modes 2, 3, and 4, except where all MSIVs are closed

and deactivated (i.e., in Modes 2, 3, and 4, TS 3.7.1.5 is applicable

only if the MSIVs are open). Also, the Action Statement for the

Limiting Condition for Operation 3.7.1.5 has been revised using the

guidance of the Improved Standard Technical Specifications for

Westinghouse plants (NUREG-1431). The amendment also deletes a license

requirement to submit responses to and to implement requirements of

Generic Letter 83-28, because the requirement has been completed.

Generic Letter 83-28 pertains to the Salem anticipated transient

without scram event. In addition, the amendment incorporates TS Bases

submitted by Northeast Nuclear Energy Company by letters dated June 20,

1995, February 5, 1996, and March 21 and 26, 1997. Since all four Bases

changes affect Section B 3/4.7 of the TS, the NRC staff is using them

in a group to avoid errors in revising the TS.

Date of issuance: April 10, 1997

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

within 60 days.

Amendment No.: 136

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

Technical Specifications and License Condition.

Date of initial notice in Federal Register: August 2, 1995 (61 FR

39445) and February 28, 1996 (61 FR 7555)The August 30, 1995, letter

provided clarifying information that did not change the scope of the

June 20, 1995, application and the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 10, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

[[Page 19843]]

Norwich, Connecticut and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, Connecticut 06385

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-

336, and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2,

and 3, New London, Connecticut

Date of application for amendments: February 3, 1997

Brief description of amendments: The amendments revise Section 6,

``Administrative Controls,'' of the Millstone Unit Nos. 1, 2, and 3

Technical Specifications to reflect organizational changes that have

been implemented in the Nuclear Division.

Date of issuance: April 10, 1997

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

60 days.

Amendment Nos.: 99, 206, and 135

Facility Operating License Nos. DPR-21, DPR-65, and NPF-49:

Amendments revised the Technical Specifications.

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

FR 8800) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 10, 1997.No significant

hazards consideration comments received: No

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

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

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

Pennsylvania

Date of application for amendments: November 18, 1996

Brief description of amendments: These amendments change the

Technical Specifications for Susquehanna Steam Electric Station (SSES),

Units 1 and 2 by increasing the maximum isolation times for reactor

core isolation cooling inboard warm-up line isolation valves from 3

seconds to 12 seconds, high pressure core injection inboard warm-up

line siolation valves from 3 seconds to 6 seconds and reactor

recirculation process sample line isolation valves from 2 seconds to 9

seconds.

Date of issuance: April 7, 1997

Effective date: Both units, as of date of issuance, to be

implemented within 30 days.

Amendment Nos.: 164 and 135

Facility Operating License Nos. NPF-14 and NPF-22: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 15, 1997 (61 FR

2191) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 7, 1997. No significant

hazards consideration comments received: No

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

Pennsylvania Power and Light Company, Docket No. 50-388,

Susquehanna Steam Electric Station, Unit 2, Luzerne County,

Pennsylvania

Date of application for amendment: March 17, 1997

Brief description of amendment: The amendment modifies the Design

Features Section 5.3.1 of the Technical Specifications to reflect the

Atrium-10 design and would include a Siemens Power Corporation topical

report in Section 6.9.3.2 to reflect mechanical design criteria for

this fuel. This change would allow this fuel to be loaded into the core

only under Operational Condition 5 (refueling) and does not permit

startup or power operation using the Atrium-10 fuel.

Date of issuance: April 9, 1997

Effective date: As of date of issuance to be implemented within 30

days.

Amendment No.: 136

Facility Operating License No. NPF-22: This amendment revised the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes (62 FR 14167) March 25, 1997.

That notice provided an opportunity to submit comments on the

Commission's proposed no significant hazards consideration

determination. No comments have been received. The notice also provided

for an opportunity to request a hearing by April 24, 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 are contained in a Safety

Evaluation dated April 9, 1997.

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

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

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366,

Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,

Georgia

Date of application for amendments: September 19, 1996, as

supplemented December 17, 1996, January 23 and 31, March 21 and April

4, 1997

Brief description of amendments: The amendments revise the

surveillance requirements addressing the reactor vessel pressure and

temperature limits.

Date of issuance: April 4, 1997

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

30 days.

Amendment Nos.: 206 and 147

Facility Operating

Local Public- Document -Room locations: ments revised the Technical

Specifications.

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

128) The December 17, 1996, January 23 and 31, March 21, 1997, and

April 4, 1997, letters provided 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 April 4, 1997.No significant hazards

consideration comments received: No

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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: October 18, 1996 as

supplemented March 12, March 17, April 4, and April 9, 1997 (TS 96-05)

Brief description of amendments: The amendments change the

Technical Specifications (TS) by revising TS 3/4.4.5 and 3.4.6.2 and

associated Bases to permanently incorporate requirements for steam

generator tube inspections and repair in the Sequoyah Nuclear Plant,

Units 1 and 2 TS.

Date of issuance: April 9, 1997

Effective date: As of the date of issuance to be implemented no

later than 45 days of its issuance.

Amendment Nos.: 222 and 213

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

revise the technical specifications and license conditions.

[[Page 19844]]

Date of initial notice in Federal Register: February 11, 1997 (62

FR 6276) The March 12, March 17, April 4, and April 9, 1997, letters

provided clarifying information that did not change the scope of the

October 18, 1996, application and the initial proposed no significant

hazards consideration determination.The Commission's related evaluation

of the amendment is contained in a Safety Evaluation dated April 9,

1997.No significant hazards consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

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 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 May 23, 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. 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

[[Page 19845]]

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

the contentions whic

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