Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from November 6, 1998, through November 19, 1998.

The last biweekly notice was published on November 18, 1998 (63 FR

64106).

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period.

[[Page 66591]]

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 and

Directives Branch, Division of Administration 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 January 4, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

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

by the above date. 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.

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,

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

Date of amendment request: November 11, 1998.

[[Page 66592]]

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) to correct Surveillance

Requirements (SRs) 3.6.11.6 and 3.6.11.7 and the associated Bases.

These SRs currently are incorrect and do not reflect the Containment

Pressure Control System (CPCS) as designed. Therefore, the proposed

amendments would only revise the SRs; no change to the CPCS design is

involved.

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:

First Standard

Implementation of this amendment would not involve a significant

increase in the probability or consequences of an accident

previously evaluated. Approval of this amendment will have no

significant effect on accident probabilities or consequences.

The CPCS is not an accident initiating system; therefore, there

will be no impact on any accident probabilities by the approval of

this amendment. The design of the CPCS is not being modified by this

proposed amendment. The amendment merely aligns [TS] surveillance

requirements with the existing design and function of the system.

Therefore, there will be no impact on any accident consequences.

Second Standard

Implementation of this amendment would not create the

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

previously evaluated. No new accident causal mechanisms are created

as a result of NRC approval of this amendment request. No changes

are being made to the plant which will introduce any new accident

causal mechanisms. This amendment request does not impact any plant

systems that are accident initiators, since the CPCS is an accident

mitigating system.

Third Standard

Implementation of this amendment would not involve a significant

reduction in a margin of safety. Margin of safety is related to the

confidence in the ability of the fission product barriers to perform

their design functions during and following an accident situation.

These barriers include the fuel cladding, the reactor coolant

system, and the containment system. The performance of these fission

product barriers will not be impacted by implementation of this

proposed amendment. The CPCS is already capable of performing as

designed. No safety margins will be impacted.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Mr. Paul R. Newton, Legal Department

(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,

North Carolina.

NRC Project Director: Herbert N. Berkow.

Duke Energy Corporation, 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: October 15, 1998.

Description of amendment request: The proposed amendments would

revise the pressure-temperature limits in the Technical Specifications

for Units 1, 2, and 3. The proposed amendments would revise the heatup,

cooldown, and inservice test limitations for the reactor coolant system

of each unit to a maximum of 26 effective full-power years. The

proposed amendments would also revise the Technical Specification for

low temperature overpressure protection to reflect the revised

pressure-temperature limits of the reactor vessels.

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. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

NO.

Each accident analysis addressed in the Oconee UFSAR [Updated

Final Safety Analysis Report] has been examined with respect to the

changes to the Reactor Pressure Vessel (RPV) pressure-temperature

limit curves and related Low Temperature Overpressure settings. The

probability of any design basis accident (DBA) is not affected by

this change, nor are the consequences of a DBA affected by this

change. The revised pressure-temperature limits, which were

developed based on NRC approved methodology or ASME Code [American

Society of Mechanical Engineers Boiler and Pressure Vessel Code]

Case N-514 as described in the Technical Justification, are not

considered to be an initiator or contributor to any accident

analysis addressed in the Oconee UFSAR. The added requirement to

deactivate one pressurizer heater bank during low temperature

operation does not significantly change the probability or

consequence of any accident previously analyzed. No existing

Technical Specification requirements are being deleted with this

revision.

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

from the accident previously evaluated?

NO.

This license amendment revises Oconee RPV pressure-temperature

limits. The revised pressure-temperature limits were developed based

on NRC approved methodology or ASME Code Case N-514 as described in

the Technical Justification. Operation of Oconee in accordance with

these proposed new Technial Specifications will not create any

failure modes not bounded by previously evaluated accidents.

Consequently, this change will not create the possibility of a new

or different accident from any accident previously evaluated.

C. Involve a significant reduction in a margin of safety?

NO.

This license amendment revises Oconee RPV pressure-temperature

limits. The revised pressure-temperature limits were developed based

on NRC approved methodology or ASME Code Case N-514 as described in

the Technical Justification. The purpose of this license amendment

is to assure that sufficient operating margin to safety is

maintained in the operation of the Oconee reactor pressure vessels

by establishing new, more limiting pressure-temperature limit curves

and adding the requirement to deactivate one pressurizer heater

bank. No plant safety limits, set points, or design parameters are

adversely affected. The fuel, fuel cladding, and Reactor Coolant

System are not impacted. Therefore, there will be no significant

reduction in any margin of safety.

Duke [Duke Energy Corporation] has concluded based on this

information 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.

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

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

NRC Project Director: Herbert N. Berkow. Duquesne Light Company, et

al., Docket No. 50-334, Beaver Valley Power Station, Unit No. 1,

Shippingport, Pennsylvania

Date of amendment request: November 11, 1998.

Description of amendment request: The proposed amendment would

modify License Condition 2.C(9) to allow, on a one time only basis, an

extension to the steam generator inspection interval of technical

specification surveillance 4.4.5.3.b. This

[[Page 66593]]

would allow the steam generator inspection interval to coincide with

the 13th refueling outage or the end of 500 effective full power days,

whichever occurs sooner.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change is temporary and allows a one time extension

of specific surveillance requirements for Cycle 13 to allow

surveillance testing to coincide with the 13th (1R13) refueling

outage. The proposed surveillance interval extension will not cause

a significant reduction in system reliability nor affect the ability

of a system to perform its design function. Current monitoring of

plant conditions and the surveillance monitoring required during

normal plant operation will be performed as usual to assure

conformance with technical specification operability requirements.

The technical specification steam generator tube inspection is

intended to prevent the Steam Generator Tube Rupture analyzed in

[Updated Final Safety Analysis Report] UFSAR Section 14.2.4 by

maintenance of the integrity of the primary to secondary coolant

boundary represented by steam generator tubes. The process by which

this integrity is maintained is inspection of steam generator tubes

at prescribed intervals, and the removal of defective tubes from

service. Inspection intervals are based on preventing corrosion

growth from exceeding tube structural limits, thereby preventing

tube failure. The 1997 steam generator inspection characterized

existing steam generator tube degradation, and degraded tubes were

removed from service at that time. Degradation growth rates were

evaluated for the next operating interval and it was determined that

the steam generator tube structural integrity is maintained.

Degradation of steam generator tubes was prevented during the

extended outage by a carefully controlled, corrosion prevention

program.

The proposed change does not affect the UFSAR and is consistent

with changes granted for other plants. The surveillance extension

does not involve a change to plant equipment and does not affect the

performance of plant equipment used to mitigate an accident. This

change, therefore, does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

Extending the surveillance interval for the performance of

specific inspections will not create the possibility of any new or

different kind of accidents. No change is required to any system

configurations, plant equipment or analyses.

Steam generator tube inspections determine tube integrity and

provide reasonable assurance that a tube rupture or primary to

secondary leak will not occur. Accidents involving steam generator

tube rupture are analyzed in UFSAR Section 14.2.4, ``Steam Generator

Tube Rupture.'' The only type of accident that can be postulated

from extending the steam generator inspection interval would be a

tube leak or rupture which are analyzed in the UFSAR. No new failure

modes are created by the surveillance extension. Therefore, this

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

accident from any accident previously evaluated.

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

of safety?

Surveillance interval extensions will not impact any plant

safety analyses since the assumptions used will remain unchanged.

The safety limits assumed in the accident analyses and the design

function of the equipment required to mitigate the consequences of

any postulated accidents will not be changed since only the

surveillance interval is being extended. Based on engineering

judgement, extending the surveillance interval for the performance

of these specific inspections does not involve a significant

reduction in the margin of safety derived from the required

surveillances.

The margin of safety depends upon maintenance of specific

operating parameters within design limits. In the case of steam

generators, that margin is maintained through assurance of tube

integrity as the primary to secondary boundary. Assurance of tube

integrity is provided through periodic in-service inspection of

tubes and removal of defective tubes from service. Additional margin

is provided through protection from possible consequences of steam

generator tube failure by mitigation systems. Radiation monitors

provide a detection capability of primary to secondary leakage to

enable a prompt response. Maintenance of the steam generator water

chemistry in accordance with [Electric Power Research Institute]

EPRI guidelines provides additional margin of safety. Therefore, the

plant will be maintained within the analyzed limits and the proposed

extension will not significantly reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: Robert A. Capra Entergy Operations, Inc.,

Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County,

Arkansas

Date of amendment request: June 30, 1998.

Description of amendment request: The proposed change modifies the

Engineered Safety Features Actuation System (ESFAS) portion of the

Arkansas Nuclear One, Unit-2 (ANO-2) Plant Protection System (PPS).

This modification is designed to defeat the backup power supply for the

auctioneered power sources for channel A and D Reactor Protective

System (RPS) and ESFAS bistables, and to provide selective logic for

Emergency Feedwater Actuation Signals and Main Steam Isolation Signals.

This will ensure that ESFAS will have the redundancy and independence

sufficient to assure that (1) no single failure results in loss of the

protection function with a channel in indefinite bypass, and (2)

removal from service of any component or channel does not result in

loss of the required minimum redundancy required by the ANO-2 Technical

Specifications (TSs). The proposed modification to the ANO-2 PPS has

been determined to involve an Unreviewed Safety Question in accordance

with 10 CFR 50.59(a)(2).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

An evaluation of the proposed change has been performed in

accordance with 10 CFR 50.91(a)(1) regarding no significant hazards

considerations using the standards in 10 CFR 50.92(c). A discussion

of these standards as they relate to this amendment request follows:

Criterion 1--Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The ANO-2 Plant Protection System (PPS) includes the electrical

and mechanical devices and circuitry (from sensors to actuation

device input terminals) involved in generating signals associated

with the two protective functions, Engineered Safety Feature

Actuation System (ESFAS) and Reactor Protective System (RPS). The

RPS is that portion of the PPS which generates signals that actuate

a reactor trip. The ESFAS is that portion of the PPS which generates

signals that actuate Engineered Safety Features (ESF) to mitigate

the consequences of an accident.

The ANO-2 Safety Analysis Report (SAR) section 15.1.31 ``Loss Of

One DC System'' analyzes failure of a DC bus (FODCB) as initiator

and its causes. The causes for the FODCB are DC leg to leg fault in

the bus or in the power distribution circuit from the battery. Since

the proposed change has no impact on the accident initiator, the

frequency of occurrence is not changed. In order for the FODCB as a

single failure with an accident to de-energize two [Vital Instrument

Buses (]VIBs[)], the FODCB would have to occur prior to the safety

bus

[[Page 66594]]

energization by offsite bus fast transfer or prior to safety bus

energization by the emergency diesel generator (EDG). The potential

for de-energization of one pair of VIBs is, therefore, limited to

time from initiation of the accident to time for safety bus response

to the secondary plant and Reactor Protective System trips.

The effects of the FODCB are being revised to assume a secondary

plant trip that results in de-energization of one power division.

The existing analysis conclusions remain unchanged. The accident

analysis is being revised to include de-energization of a pair of

vital AC instrument channels. De-energization of two vital AC

sources has not been previously documented as a design bases event.

Auctioneered bistable power supplies for Plant Protection System

(PPS) channels A and D are being modified to a single power source

for each of these two channels. Single channel trips will result for

all PPS functions in channels A or D for loss of its single channel

bistable power source. The PPS channels B and C auctioneered power

supplies remain unchanged to maintain Recirculation Actuation Signal

(RAS) response to a FODCB.

Regarding PPS measurement channels with increasing signal

setpoints, de-energization of a single power supply either results

in failure of a measurement channel (B or C) to a non-tripped state

or in failure of a measurement channel (A or D) to a tripped state.

Neither single channel failure scenario impacts accident initiation

or mitigation. For PPS measurement channels with decreasing signal

setpoints the single channel de-energization events result in

failure of a single affected measurement channel to a tripped state.

The PPS two out of three logic design with a channel bypassed

ensures operability with a single channel failure. Neither condition

impacts accident frequency or consequences.

With the exception of Recirculation Actuation Signal (RAS) and

Emergency Feedwater Actuation Signal (EFAS), a FODCB results in an

automatic ESFAS initiation for those functions with decreasing

signal setpoints. For other ESFAS functions with a decreasing

signal, channels A and C or channels B and D fail to the tripped

state. For those functions with an increasing signal setpoint

(including EFAS), a FODCB results in a single channel failing not

tripped, one channel tripping, and two channels remaining

functional. System level functions remain operable with either a one

out of two logic (no channels bypassed) or a one out of one logic

(with a channel bypassed).

Interposing relay actuation logic has changed from single trip

path to selective trip path logic. This change insures emergency

feedwater (EFW) discharge valves will receive an automatic open or

close demand based on steam generator level and pressure demands.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

Criterion 2--Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

In response to de-energization of a pair of Vital Instrument

Buses (VIBs), those ESFAS functions with increasing signal

setpoints, as a minimum, remain functional with one out of one

logic. One channel trips, one channel does not trip, and two

channels remain functional. One of the functional channels may be

bypassed without impact on operability. The trip response of those

ESFAS functions with decreasing signal to trip setpoints remains

unchanged.

EFAS coincidence logic to close the EFW discharge valves

requires three out of four channels to be in a non-tripped state.

With a FODCB one channel is tripped, one channel is not tripped, and

two channels are functional. The close logic becomes two out of two

with a FODCB.

By defeating the auctioneered bistable power sources for PPS

channel A and D bistables, PPS measurement channel A or D will fail

to its tripped state. This change ensures no more than one channel

(B or C) fails to a non-tripped state for the FODCB.

With selective logic EFAS pump discharge valves will receive

control signals to initiate emergency feedwater and to terminate

emergency feedwater flow by open and close demands generated

independent of the 120 Volt channel pair de-energization.

The existing ANO-2 Failure Modes and Effects Analysis does not

document failure of a pair of vital instrument AC channels. Neither

the 120 Volts AC nor the 125 Volt DC system single failure analysis

assumes failure of two channels of 120 Volts AC. Even though the

failure of either pair of VIBs caused by a FODCB is not a result of

the proposed change, the SAR change will address the potential for

de-energization of a pair of instrument buses. The ANO-2 SAR will be

updated to reflect the documentation and modification of the PPS

design to ensure safe plant response.

Even though the plant response to FODCB is being modified, the

proposed ANO-2 PPS design resolution does not create the possibility

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

in the SAR. The PPS will have the redundancy and independence

sufficient to assure that (1) no single failure results in loss of

the protection function, and (2) removal from service of any

component or channel does not result in loss of the required minimum

redundancy required by the TS. PPS will also meet the single failure

criterion of IEEE 279-1971 to the extent that any single failure

within the system does not prevent proper protective action at the

system level and no single failure will defeat more than one of the

four protective channels associated with any one trip function.

Criterion 3--Does Not Involve a Significant Reduction in the

Margin of Safety.

Technical Specification Bases 3/4.3.1 & 3/4.3.2 assure

sufficient PPS redundancy is maintained to permit a channel to be

bypassed. Under the current design, a FODCB will result in reduction

of margin by decreasing the number of functional channels to less

than two. However, with the proposed modification removal from

service of any component or channel for indefinite bypass will not

result in loss of the minimum redundancy required by the TS. This

activity will restore the margin by ensuring ESFAS required

functions remain capable of automatic actuation with a FODCB.

Therefore, this change does not involve a significant reduction

in the margin of safety.

Based upon the reasoning presented above and the previous

discussion of the amendment request, Entergy Operations has

determined that even though the proposed PPS design description

results in an accident or malfunction of a different type, the

requested change does not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

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

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: October 29, 1998.

Description of amendment request: The proposed amendment would

revise the terminology used in the St. Lucie Plant Technical

Specifications (TS) relative to the implementation and automatic

removal of certain reactor protection system trip bypasses to ensure

that the meaning of explicit terms used in the TS are consistent with

the intent of the stated requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments are administrative in nature, and do not

change the function or the setpoints of the RPS trip bypass

features. The revisions simply make corrections to the Notation of

TS Tables 2.2-1 and 3.3-1 to ensure that the meaning of explicit

terms used in the Notes is consistent with the intent of the stated

requirements based on the St. Lucie plant design. The proposed

technical specification changes do not involve accident initiators,

do not change the configuration or method of operation of any plant

equipment that is used to mitigate

[[Page 66595]]

the consequences of an accident, and do not alter any conditions

assumed in the plant accident analyses. Therefore, operation of

either facility in accordance with its proposed amendment would not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendments are administrative in nature and will

not change the physical plant or the modes of plant operation

defined in the facility operating licenses. The changes do not

involve the addition or modification of equipment nor do they alter

the design or operation of plant systems. Therefore, operation of

either facility in accordance with its proposed amendment would not

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

any accident previously evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendments are administrative in nature and do not

change the function or the setpoints of the RPS trip bypass

features. The revisions simply make corrections to the Notation of

TS Tables 2.2-1 and 3.3-1 to ensure that the meaning of explicit

terms used in the Notes is consistent with the intent of the stated

requirements based on the St. Lucie plant design. The proposed

changes do not alter the basis for any technical specification that

is related to the establishment of, or the maintenance of, a nuclear

safety margin. Therefore, operation of either facility in accordance

with its proposed amendment would not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration. This

notice is intended to replace an exigent notice of consideration of

issuance of amendment for St. Lucie Unit 1, previously published as

exigent TS amendments for both St. Lucie Units 1 and 2 in the Federal

Register (63 FR 59809). The amendment request for St. Lucie Unit 2 will

continue to be considered as an exigent amendment as noticed in the

Federal Register (63 FR 59809).

Local Public Document Room location: Indian River Junior College

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Project Director: Frederick J. Hebdon.

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

Generating Station, Ocean County, New Jersey

Date of amendment request: November 10, 1998.

Description of amendment request: The proposed Technical

Specification (TS) change would remove the restriction on the sale or

lease of property within the exclusion area and replace the restriction

with a requirement to retain complete authority to determine and

maintain sufficient control of all activities including the authority

to exclude or remove personnel and property within the minimum

exclusion distance. A TS Bases page for the proposed change is

included. Also included are clarifications and administrative changes

which (1) clarify TS definition 1.38 to become ``Site Boundry'' from

the current term ``Exclusion Area'' to be consistent with 10 CFR

20.1003 definition for Site Boundry and the 10 CFR 100.3 definition of

Exclusion Area, (2) convert the one occurrence of the use of TS

definition from Exclusion Area to Site Boundry in TS 6.8.4(a)(9), and

(3) revise and update the Table of Contents for Section I Definitions.`

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. Would operation of the facility in accordance with the

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

The proposed change is administrative in nature and does not

affect the purpose, function, performance, operability or testing of

and does not make any physical or procedural changes to plant

systems, structures or components. Also, all existing technical

specification limiting conditions for operation and surveillance

requirements are retained.

[Technical Specification Change Request] TSCR 264 does not

change the size or location of the exclusion area. Since the

exclusion area size and location are not being changed and no

physical or procedural changes are being made to the plant,

radiological consequences in the exclusion area are not affected by

this TSCR.

This change addresses the existing technical specification

restriction on the sale or lease of property within the ``exclusion

area'' by ensuring that the licensee will retain at all times the

complete authority to determine and maintain sufficient control of

all activities through ownership, easement, contract and/or other

legal instruments on property within the minimum exclusion distance

including the authority to exclude or remove personnel and property

within the minimum exclusion distance.

Therefore, since no physical or procedural changes are being

made to existing plant systems, structures or components and since

the proposed change requires the licensee to retain complete

authority and sufficient control of all activities in the exclusion

area, operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. Would operation of the facility in accordance with the

proposed change create the possibility of a new or different kind of

accident from any accident previously evaluated?

The p[ro]posed change is administrative in nature and does not

affect the purpose, function, performance, operability or testing of

and does not make any physical or procedural changes to plant

systems, structures or components. Also, all existing technical

specification limiting conditions for operation and surveillance

requirements are retained.

This change addresses the existing technical specification

restriction on the sale or lease of property within the ``exclusion

area'' by ensuring that the licensee will retain at all times the

complete authority to determine and maintain sufficient control of

all activities through ownership, easement, contract and/or other

legal instruments on property within the minimum exclusion distance

including the authority to exclude or remove personnel and property

within the minimum exclusion distance.

Therefore, since no physical or procedural changes are being

made to existing plant systems, structures or components and since

the proposed change requires the licensee to retain complete

authority and sufficient control of all activities in the exclusion

area, operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3. Would operation of the facility in accordance with the

proposed change involve a significant reduction in a margin of

safety?

The p[ro]posed change is administrative in nature and does not

affect the purpose, function, performance, operability or testing of

and does not make any physical or procedural changes to plant

systems, structures or components. Also, all existing technical

specification limiting conditions for operation and surveillance

requirements are retained.

This change addresses the existing technical specification

restriction on the sale or lease of property within the ``exclusion

area'' by ensuring that the licensee will retain at all times the

complete authority to determine and maintain sufficient control of

all activities through ownership, easement, contract and/or other

legal instruments on property within the minimum exclusion distance

including the authority to exclude or remove personnel and property

within the minimum exclusion distance.

Therefore, since no physical or procedural changes are being

made to existing plant

[[Page 66596]]

systems, structures or components and since the proposed change

requires the licensee to retain complete authority and sufficient

control of all activities in the exclusion area, operation of the

facility in accordance with the proposed amendment will not involve

a significant reduction in 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: Ocean County Library,

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

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

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

NRC Project Director: Cecil O. Thomas.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2 (NMP2), Oswego County, New York

Date of amendment request: October 16, 1998.

Description of amendment request: The proposed amendment would make

the following revisions to Technical Specifications (TSs) 3/4.7.1.1:

(1) Ensure that four service water (SW) pumps are operating with the

divisional cross connect valves open during Operational Condition 1, 2

and 3 (current TS requires two SW pumps associated with one loop to be

operating); (2) Increase the number of division 1 and 2 heaters

required to be operable from 7 per division per intake to 14 per

division per intake; (3) The actions necessary for having less than the

required equipment is being revised to reflect the new limits for SW

equipment; and (4) SW supply header discharge water temperature is

being increased from 81 to 82 deg.F. TS 3.7.1.2, Table 3.3.9-1, and

Table 4.3.9.1-1 are revised to add ``when handling irradiated fuel in

the secondary containment'' to the applicability section. Table 3.3.9-1

is being revised to decrease the temperature at which the Intake

Deicing Heaters are required to be in service from 39 to 38 degrees F.

TS 3.7.1.2 proposed change is to specify that the necessary portions of

the SW system needed to support equipment required to be operable shall

be operable; the Action Section proposed revision reflects this change.

TS 4.7.1.2.1 surveillance requirement proposed change is to increase

the flow rate of SW pumps from 6500 GPM to 9000 GPM and to change the

SW pumps pressure from 80 psi discharge pressure to 70 psi differential

pressure; TS 4.7.1.2.2 is being revised to decrease the intake tunnel

water temperature from 39 to 38 degrees F. The surveillance for the

Intake Deicing Heaters is being changed to reflect the increase in the

number of heaters required. The title of ``Plant Service Water System''

is being changed to ``Service Water System.''

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 operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The SW System is a once-through system which supplies water from

Lake Ontario to various essential and non-essential components, as

required, during normal plant operation and shutdown conditions. The

System is designed with suitable redundancy to provide a reliable

source of cooling water for the removal of heat from essential plant

components, including the RHR [residual heat removal] heat

exchangers, the EDGs [emergency diesel generators], and room coolers

for ECCS [emergency core cooling system] equipment, which are

required for safe reactor shutdown following a LOCA.

LCO 3.7.1.1 and LCO 3.7.1.2 each currently requires two

independent SW System loops to be operable, with one of the loops in

operation. The current LCOs do not provide adequate guidance

regarding the minimum number of operating pumps. NMPC [Niagara

Mohawk Power Corporation] proposes to revise LCO 3.7.1.1 and its

associated Actions and SRs to provide assurance that four SW pumps

are operable and are operating within acceptable system parameters,

with the divisional cross-connect valves open, during Operational

Conditions 1, 2, and 3 to meet the limiting LOCA analysis

assumptions.

TS Section 3/4.7.1 currently specifies a maximum SW supply

header discharge water temperature of 81 degrees F and a limiting

temperature for Intake Deicing Heater ystem operability (intake

water) temperature of 39 degrees F. In addition, TS Table 3.3.9-1,

Action 144, requires the Intake Deicing Heater System heaters to be

placed in service when the Lake Ontario water temperature reaches 39

degrees F. NMPC proposes to revise Action 144 of TS Table 3.3.9-1

and TS LCO 3.7.1.1, including its associated Actions and SRs

[surveillance requirements], to increase the supply header discharge

water temperature to its analytical limit of 82 degrees F and reduce

the limiting temperature for the Intake Deicing Heater System Action

and operability requirements to 38 degrees F.

Appropriate changes to LCO 3.7.1.2 and its associated Actions

and SRs are also proposed in order to assure consistency with the SW

System analyses assumptions during shutdown conditions. The current

LCO Actions do not account for the varying flows and heat loads that

may be required for various plant shutdown conditions. The revision

to the Applicability for LCO 3.7.1.2 and TS Tables 3.3.9-1 and

4.3.9.1-1 will assure that the SW System is operable during periods

when irradiated fuel is being handled in the secondary containment

and essential loads cooled by the SW System are required to be

operable (e.g., EDG). A footnote has been added to define

Operational Condition * and is consistent with similar footnotes in

the TSs. The proposed changes will assure that the necessary ortions

of the SW System and the necessary Divisions of the Intake Deicing

Heater System heaters are operable that are supporting equipment

required to be operable.

It is further proposed to change the system title identified in

the Index and in TS Section 3/4.7.1, including the LCOs and SRs,

from ``Plant Service Water System'' to ``Service Water System'' to

be consistent with the NMP2 [Nine Mile Unit 2] UFSAR [Updated Final

Safety Analysis Report].

The changes do not involve any physical alteration of the plant,

and the SW System will remain capable of providing sufficient

cooling flow for the essential cooling loads during plant operation

and also during plant shutdown. The changes will have no impact on

the design or function of the SW System and its components, thus

assuring that the characteristics and functional performance are

maintained consistent with the event precursors and the conditions

and assumptions of the current design basis accident and transient

analyses. The changes to the LCO AOTs [allowed outage times] are

either consistent with or are more conservative than the current

AOTs. Based on the above, adequate assurance is provided that the

probability of event initiation will remain as previously analyzed.

Maintaining four pumps operating within acceptable system

parameters, with the divisional cross connect valves open, during

Operational Conditions 1, 2, and 3 provides assurance that the

essential functions supported by the SW System are maintained.

Particularly, adequate SW flow assures that the primary and

secondary containments can perform their intended functions of

limiting the release of radioactive materials to the environment

following a LOCA. The small (1 degree F) change in the SW supply

header discharge water (UHS) temperature and Intake Deicing Heater

System actuation temperature maintain the current design basis for

the UHS and SW Systems such that there will be no impact on the LOCA

analyses assumptions or conclusions. The proposed changes to the SW

System TSs do not adversely affect the capability of plant systems,

structures, and components to respond to any accident in Operational

Conditions 4, 5, and *. As a result, there will be no degradation of

the primary or secondary containment or any other fission product

barriers which could increase the radiological consequences of an

accident. In addition, other essential accident mitigation equipment

supported by the SW System will not be adversely impacted. It is,

therefore,

[[Page 66597]]

concluded that operation of NMP2, in accordance with the proposed

amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated. The

operation of Nine Mile Point Unit 2, in accordance with the proposed

amendment, will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The changes do not result in any hardware changes or physical

alteration of the plant which could introduce new equipment failure

modes, and there will be no impact on the design or function of the

SW System or its components. The primary and secondary containment

post-LOCA responses remain within previously assessed limits of

temperature and pressure. Furthermore, adequate cooling flow is

assured during plant operation and also during shutdown conditions

such that essential systems and components remain within their

applicable design limits. It is, therefore, concluded that no

requirements are eliminated or new requirements imposed which could

affect equipment or plant operation such that new credible accidents

are introduced. Accordingly, operation of NMP2, in accordance with

the proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The changes provide assurance that the SW System will remain

capable of providing sufficient cooling flow for the essential

cooling loads during plant operation and also during plant shutdown

such that essential systems and components remain within their

applicable design limits. The changes will have no impact on the

design or function of the SW System and its components, thus

assuring that the characteristics and functional performance are

maintained consistent with the conditions and assumptions of the

current design basis accident and transient analyses. Maintaining

four pumps operating within acceptable system parameters, with the

divisional cross connect valves open, during Operational Conditions

1, 2, and 3 provides assurance that post-LOCA radioactive releases

are maintained within 10 CFR 100 limits. The small (1 degree F)

change in the SW supply header discharge water (UHS) temperature and

the limiting temperature for the Intake Deicing Heater System Action

and operability requirements maintains the current design basis for

the UHS and SW Systems such that there will be no impact on the LOCA

analyses assumptions or conclusions.

These changes will not result in a reduction in margin to the

System analytical limits. Furthermore, maintaining the intake bar

surface temperature at least 1 degree F above freezing provides an

adequate margin to prevent the adherence of ice, and provides

assurance that sufficient flow area is always heated such that the

SW System will remain capable of providing adequate cooling flow in

the event of a LOCA. Similarly, maintaining the required SW System

flow and temperature during Operational Conditions 4, 5, and * will

assure that the associated equipment is operable such that

radioactive releases are maintained within 10 CFR 100 limits. It is,

therefore, concluded that the changes do not eliminate any

requirements, impose any new requirements, or alter any physical

parameters which significantly reduce the margin to an acceptance

limit or adversely affect the margins associated with the fission

product barriers as established by the design basis accident and

transient analyses. Accordingly, operation of NMP2, in accordance

with the proposed amendment, will not involve a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Reference and Documents

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

York 13126.

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

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

NRC Project Director: S. Singh Bajwa.

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

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

Connecticut

Date of amendment request: September 28, 1998.

Description of amendment request: The proposed amendment would

change Technical Specifications 3.3.2.1, ``Instrumentation--Engineered

Safety Features Actuation System''; 3.4.6.2, ``Reactor Coolant System--

Reactor Coolant System Leakage''; 3.4.8, ``Reactor Coolant System--

Specific Activity''; 3.6.2.1, ``Containment Systems--Depressurization

and Cooling Systems Containment Spray and Cooling Systems''; 3.6.5.1,

``Containment Systems--Secondary Containment Enclosure Building

Filtration System''; 3.7.6.1, ``Plant Systems--Control Room Emergency

Ventilation System''; and 3.9.15, ``Refueling Operations--Storage Pool

Area Ventilation System--Fuel Storage.'' Information would also be

added to the Bases of the associated Technical Specifications to

address the proposed changes.

The proposed amendment would also revise the Operating License DPR-

65 by incorporating a change to the Millstone Unit No. 2 Final Safety

Analysis Report (FSAR). The change to the FSAR is associated with the

revised main steamline break analyses, new determination of the

radiological consequences of a main steamline break, and a revised

determination of the radiological consequences of the design basis

loss-of-coolant accidents (LOCAs).

The proposed changes to the main steamline break analysis, as

described in the FSAR, are based on the revised Siemens Power

Corporation steamline break methodology. The report describing the

revised methodology was submitted by Siemens Power Corporation to the

NRC for approval in a letter dated June 30, 1998. The revised

methodology was used to perform the Millstone Unit No. 2 plant-specific

analysis for post-scram main steamline break. This plant-specific

analysis was submitted by NNECO in a letter dated August 12, 1998,

which proposed to change the list of documents in the Technical

Specifications that describe the analytical methods used to determine

the core operating limits. The proposed changes contained in this

letter assume approval of the previously submitted revised Siemens

Power Corporation steamline break methodology, and the changes to the

list of documents in the Millstone Unit No. 2 Technical Specifications

that describe the analytical methods used to determine the core

operating limits.

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:

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

changes and has concluded that they do not involve a significant

hazards consideration (SHC). The basis for this conclusion is that

the three criteria of 10CFR50.92(c) are not compromised. The

proposed changes do not involve an SHC because the changes would

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Analyses Changes

The main steam line break analyses and the determinations of the

radiological consequences of the main steam line break and loss of

coolant accident have been revised. A brief summary of the

significant changes to the main steam line break analyses and the

radiological consequences of the main steam line break and loss of

coolant accident is presented below.

1. The limited fuel failure following a main steam line break

outside containment results in an increase in the calculated

radiological consequences both off-site and in the control room. To

limit the consequences of a main steam line break outside

containment, the

[[Page 66598]]

Technical Specification allowed steam generator tube leakage will be

reduced to 0.035 gpm [gallons per minute] per steam generator.

2. Credit will now be taken for iodine removal from the

containment atmosphere by the Containment Spray System (CSS). The

use of the CSS for iodine removal has not been previously approved

by the NRC.

3. The proposed increase to the allowable control room in-

leakage will provide additional operational flexibility to address

expected minor system degradation over time. The increase in the

allowable control room in-leakage will result in an increase in the

calculated dose to the Control Room Operators.

4. The addition of the dose consequences from containment sump

backleakage to the Refueling Water Storage Tank (RWST) has been

included in the off-site and control room loss of coolant accident

(LOCA) analyses increases the consequences of previously evaluated

accidents.

The containment sump backleakage into the RWST results in sump

water entering the RWST when the RWST is at its minimum level. The

RWST will become a radioactive source and contribute a shine dose to

the surrounding areas. The increase in dose rates onsite will not

prevent operators from remaining in the control room or from

accessing equipment needed to mitigate the accident.

All piping and valves associated with RWST backleakage are

located in a harsh radiation area. Backflow from the sump might

increase dose rates in the area where these components are located.

Additional dose contributions, where they occur, do not adversely

impact the environmental qualification of the vital equipment

located there. All vital equipment would continue to perform its

safety function.

5. Credit will be taken in the main steam line break analyses

for the recently installed cavitating venturis in the Auxiliary

Feedwater System. However, this will not change the amount of fuel

failure. Therefore, credit for this equipment will not impact the

radiological consequences of a main steam line break.

6. Credit will be taken for the Reactor Coolant System (RCS) low

flow reactor trip for the pre-scram inside containment main steam

line break analysis. This equipment will be qualified for the

expected containment environment following a main steam line break

inside containment and will be added to the Environmental

Qualification Master List.

7. Millstone Unit No. 1 design basis accidents, loss of coolant

and main steam line break, will no longer be evaluated for impact on

Millstone Unit No. 2 control room habitability. This credits the

decision to decommission Millstone Unit No. 1. [Footnote--B.D.

Kenyon letter to the NRC, ``Millstone Nuclear Power Station, Unit

No. 1 Certification of Permanent Cessation of Power Operations and

that Fuel Has Been Permanently Removed from the Reactor,'' dated

July 21, 1998.]

The revised main steam line break analyses and the revised

determinations of the radiological consequences of the main steam

line break and design basis LOCA analyses take credit for equipment

not previously assumed in the analyses, and for plant or equipment

operating restrictions not currently contained in the Technical

Specifications. The changes to the analyses will not adversely

affect the probability of an accident previously evaluated, but the

revised analyses results do indicate that the consequences of an

accident previously evaluated will increase. Specifically, the

following changes cause an increase in the consequences of an

accident previously evaluated.

1. The increase in allowable control room in-leakage from 100

SCFM [standard cubic feet per minute] to 130 SCFM when the Control

Room Emergency Ventilation System is operating in the recirculation/

filtration mode.

The dose to the Control Room Operators from a Millstone Unit No.

2 LOCA increased from 9.25 to 25.8 rem to the thyroid and from 0.205

to 2.29 rem to the skin. The dose to the whole body decreased. (Both

low wind speed and high wind speed release conditions were analyzed.

The low wind speed condition bounds the high wind speed condition.)

The dose to the Control Room Operators from a Millstone Unit No. 3

LOCA increased from 2.67 to 14 rem to the skin and from 0.209 to

1.484 rem to the whole body. The dose to the thyroid decreased. The

doses to the Control Room Operators from either a Millstone Unit No.

2 or Unit No. 3 LOCA remain below the GDC [General Design Criterion]

19 criteria of 30 rem thyroid, 5 rem whole body and 30 rem to the

skin.

The new calculated doses to the Millstone Unit No. 2 Control

Room Operators from a main steam line break outside containment are

29 rem thyroid, 0.03 rem whole body and 0.5 rem skin. The doses to

the Millstone Unit No. 2 Control Room Operators are below the GDC 19

criteria of 30 rem thyroid, 5 rem whole body, and 30 rem to the

skin. (Note: The dose to the Control Room Operators from a main

steam line break was not previously evaluated because fuel failure

was not predicted to occur.)

2. The limited fuel failure that is predicted in the revised

main steam line break analyses.

Previously, the radiological consequences of a main steam line

break were not determined and were not presented in the FSAR because

fuel failure was not predicted to occur. Because of the predicted

limited fuel failure for the main steam line break outside of

containment, the radiological consequences were analyzed. The

results to the Exclusion Area Boundary (EAB) are 4.8 rem thyroid and

0.06 rem whole body. The results to the Low Population Zone (LPZ)

are 2.3 rem thyroid and 0.02 rem whole body. To meet the dose

acceptance criteria to the Millstone Unit No. 2 Control Room

Operators, the maximum allowable Technical Specification primary to

secondary leak rate is being reduced to 0.035 gpm per steam

generator. The results to the Millstone Unit No. 2 Control Room

Operators are 29 rem thyroid, 0.03 rem whole body and 0.5 rem skin.

The main steam line break outside containment is the limiting

accident for the Millstone Unit No. 2 Control Room Operators.

However, the dose consequences of a main steam line break are less

than the 10CFR100 limits off-site of 300 rem thyroid and 25 rem

whole body, and the doses to the Millstone Unit No. 2 Control Room

Operators are below the GDC 19 criteria of 30 rem thyroid, 5 rem

whole body, and 30 rem to the skin.

3. Taking credit for the low RCS flow reactor trip for the pre-

scram inside containment main steam line break analysis.

Previous analyses did not credit the low RCS flow reactor trip

in a harsh environment. This credits the low flow trip in a manner

not previously reviewed by the NRC for Millstone Unit No. 2. Without

credit for this reactor trip, the predicted fuel failure for steam

line breaks inside containment would be higher.

4. Taking credit for the removal of radioactive iodine from the

containment atmosphere by containment spray.

Previous analyses did not rely on the spray function to reduce

iodine concentration in the post-accident atmosphere inside

containment. This adds a mitigation function to the CSS that has not

been previously reviewed by the NRC for Millstone Unit No. 2.

Without credit for the removal of iodine, the predicted dose

consequences following a LOCA would be higher.

5. The addition of sump backleakage to the RWST during a LOCA.

The resultant dose contribution to the LPZ from RWST backleakage

is 1.487 rem thyroid and 0.11 rem whole body. The total dose to the

LPZ from a design basis LOCA is 21.86 rem thyroid and 0.941 rem

whole body. The dose is well below the 10CFR100 limits of 300 rem

thyroid and 25 rem whole body. The dose to the EAB was not affected

because leakage into the RWST does not start until 25.45 hours post-

LOCA and the EAB is a 2-hour dose.

The resultant dose contribution to the Millstone Unit No. 2

Control Room Operators from RWST backleakage is 3.75 rem thyroid,

0.017 rem whole body and 0.296 to the skin. The total dose to the

Millstone Unit No. 2 Control Room Operators from the LOCA is 25.8

rem thyroid, 0.718 rem whole body and 2.29 rem to the skin. These

doses are below the GDC 19 limits of 30 rem thyroid and skin, and 5

rem whole body.

The analyses results meet the guidance contained in SRP

[Standard Review Plan] 15.1.5, SRP 15.6.5, and the limits of

10CFR100 and GDC 19. Therefore, there will be no significant

increase in the probability or consequences of an accident

previously evaluated.

Technical Specification Changes

Technical Specification Non-Technical Changes

The minor editorial and non-technical changes to correct

spelling (Technical Specification 3.3.2.1), modify the title of a

table column (Technical Specification 3.4.8), clarify the type of

measurement performed (Technical Specification 3.4.8), and establish

consistent terminology (Technical Specification 3.7.6.1) will not

result in any technical changes to the Millstone Unit No. 2

Technical Specifications. The proposed changes will have no adverse

effect on plant

[[Page 66599]]

operation. Therefore, there will be no significant increase in the

probability or consequences of an accident previously evaluated.

Technical Specification 3.4.6.2

The reduction in the maximum allowable value of primary to

secondary leakage per steam generator is consistent with the new

radiological assessment of the potential control room operator

exposure following a main steam line break outside of containment.

The wording change to SR [Surveillance Requirement] 4.4.6.2.1 will

clarify that the water inventory balance is used to verify

compliance with the identified and unidentified leakage limits.

Pressure boundary leakage would first show up as unidentified

leakage during performance of SR 4.4.6.2.1. Further investigation,

(plant walkdown) would be necessary to classify the unidentified

leakage as pressure boundary leakage. This is consistent with

established plant practices to detect pressure boundary leakage.

The addition of the new SR 4.4.6.2.2 will address the primary to

secondary leakage limit. The new SR will include an exception to

Technical Specification 4.0.4 that will allow the determination of

primary to secondary leakage to be deferred until after Mode 4 is

entered. Even though verification of compliance with the primary to

secondary limit will not be done prior to entering Mode 4, the limit

is still expected to be met.

The proposed changes will have no adverse effect on plant

operation. Therefore, there will be no significant increase in the

probability or consequences of an accident previously evaluated.

Technical Specification 3.4.8

The addition of the words ``of gross specific activity'' to the

Limiting Condition for Operation (LCO), Action Statements, and SR

will clarify what the E-Bar limit applies to. This is consistent

with the Technical Specification Definition (1.20) for E-Bar.

The addition of a footnote (*) to state the power history

requirements for the determination of E-Bar will ensure that the

necessary plant conditions are established prior to performing the

analysis. This will not affect the E-Bar LCO limit or the

requirement to perform the analysis. The proposed change is

consistent with NUREG--0212 and NUREG--1432.

The footnote will also specify that the provisions of

Specification 4.0.4 are not applicable. This will allow entry into

Mode 1, without determining the value of E-Bar, assuming that the

power history requirements will not be met until after Mode 1 is

entered. This will normally only apply following an extended

shutdown.

The Isotopic Analysis for Iodine (including I-131, I-133, and I-

135) sample requirement will be expanded to include the LCO

requirement for 100/E-Bar. This is consistent with the requirements

of Action Statement d. This change will expand the sampling

requirement for iodine. Minor wording changes will also be made to

be consistent with the proposed changes to the LCO wording.

The proposed changes will have no adverse effect on plant

operation. Therefore, there will be no significant increase in the

probability or consequences of an accident previously evaluated.

Technical Specification 3.6.2.1

The revised radiological assessment calculation for the design

basis accident credits iodine removal from the containment

atmosphere by the CSS. This will require a reduction in the allowed

outage time (AOT) of one containment spray train from seven days to

seventy two hours. This AOT is consistent with NUREG-0212 and NUREG-

1432. This will help ensure that plant equipment assumed in the

safety analyses will be available. This is a more restrictive change

which will have no adverse effect on plant operation. Therefore,

there will be no significant increase in the probability or

consequences of an accident previously evaluated.

Technical Specification 3.6.5.1

The value for the pressure drop across the combined HEPA [high-

efficiency particulate air] filters and charcoal adsorber banks

specified in SR 4.6.5.1.d.1 will be changed from a generic value

[less than or equal to] 6 inches water gauge) to a plant specific

value [less than or equal to] 2.6 inches water gauge). This is a

more restrictive change which will have no adverse effect on plant

operation. Therefore, there will be no significant increase in the

probability or consequences of an accident previously evaluated.

Technical Specification 3.7.6.1

The value for the pressure drop across the combined HEPA filters

and charcoal adsorber banks specified in SR 4.7.6.1.e.1 will be

changed from a generic value [less than or equal to] 6 inches water

gauge) to a plant specific value [less than or equal to] 3.4 inches

water gauge). This is a more restrictive change which will have no

adverse effect on plant operation.

SR 4.7.6.1.e.2 will be expanded to clarify that the test of the

capability of the Control Room Emergency Ventilation Trains to

switch to the recirculation mode is performed with the trains

initially operating in the normal mode and the smoke purge mode of

operation. This will not affect the requirement that the trains be

capable of switching to the recirculation mode.

The value of allowable control room air in-leakage specified in

SR 4.7.6.1.e.3 will be increased from 100 SCFM to 130 SCFM. This is

consistent with the recently revised control room radiological

analysis for the design basis accidents.

The proposed increase will provide additional operational

flexibility to address expected minor system degradation over time.

This increase is supported by the new analysis.

The proposed changes will have no adverse effect on plant

operation. Therefore, there will be no significant increase in the

probability or consequences of an accident previously evaluated.

Technical Specification 3.9.15

The value for the pressure drop across the combined HEPA filters

and charcoal adsorber banks specified in SR 4.9.15.d.1 will be

changed from a generic value [less than or equal to] 6 inches water

gauge) to a plant specific value [less than or equal to] 2.6 inches

water gauge). This is a more restrictive change which will have no

adverse effect on plant operation. Therefore, there will be no

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes have no adverse effect on how any of the

associated systems or components function to prevent or mitigate the

consequences of design basis accidents. Also, the proposed changes

have no adverse effect on any design basis accident previously

evaluated since the changes are consistent with the revised

analyses, and the appropriate acceptance criteria are met for the

revised analyses. Therefore, 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. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

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.

Also, the response of the plant and the operators following

these accidents is unaffected by the change. Therefore, the proposed

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.

Analyses Changes

The acceptance criteria for a main steam line break in the SRP

15.1.5 does not exclude the prediction of fuel failure. Instead, the

SRP requires that ``Any fuel damage calculated to occur must be of

sufficiently limited extent that the core will remain in place and

intact with no loss of core cooling.'' The limited fuel failure that

is now predicted in the revised main steam line break analyses meets

this acceptance criterion. In addition, the RCS low flow reactor

trip that is now being credited to function in a harsh environment

to limit fuel failure is already required to be operable by

Technical Specifications.

The revised dose consequences for the design basis accidents

assumes a control room in-leakage of 130 SCFM. In addition, iodine

removal by the CSS, which is already required to be operable by

Technical Specifications, is assumed. The acceptance criteria for

the dose consequences of the design basis accidents to the EAB, LPZ

and the control room personnel is met in the revised analyses.

Therefore, the revisions to the dose consequence analyses for the

design basis accidents do not involve a significant reduction in the

margin of safety.

[[Page 66600]]

Technical Specification Changes

The proposed changes will correct spelling and terminology

errors, reduce the maximum allowable primary to secondary leakage,

add a new surveillance requirement, modify surveillance requirements

for RCS specific activity, reduce the allowed outage time for a

containment spray train, reduce the allowed pressure drop across the

control room and enclosure building HEPA [high-efficiency

particulate air] filters, and increase the control room maximum

allowed in-leakage. These changes will have no adverse effect on

equipment important to safety. The equipment will continue to

function as assumed in the design basis accident analysis.

Therefore, there will be no significant reduction of the margin of

safety as defined in the Bases for the Technical Specifications

affected by these proposed changes.

The only adverse impact of the proposed changes is that the dose

consequences following an accident may increase. However, the

revised analyses show that the acceptance criteria for the accident

analyses are met. Therefore, based on the responses above, the

proposed changes are deemed safe.

The NRC has provided guidance concerning the application of

standards in 10CFR50.92 by providing certain examples (March 6,

1986, 51 FR 7751) of amendments that are considered not likely to

involve an SHC. The minor editorial and non-technical changes

proposed herein to correct reference, spelling, and terminology

errors are enveloped by example (i), a purely administrative change

to Technical Specifications. The changes proposed herein to add a

new surveillance requirement to verify primary to secondary leakage

and to reduce the allowable pressure drop across various ventilation

filters are enveloped by example (ii), a change that constitutes an

additional limitation, restriction, or control not presently

included in the Technical Specifications. All of the other changes

proposed herein are not enveloped by any specific example.

As described above, this License Amendment Request does not

impact the probability of an accident previously evaluated, does not

involve a significant increase in the consequences of an accident

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

different kind of accident from any accident previously evaluated,

and does not result in a significant reduction in a margin of

safety. Therefore, NNECO has concluded 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,

Connecticut.

NRC Project Director: William M. Dean.

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

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

Date of amendment request: October 22, 1998.

Description of amendment request: The licensee is proposing to

change Technical Specifications 3.3.2.1, ``Instrumentation--Engineered

Safety Feature Actuation System Instrumentation''; 3.4.9.3, ``Reactor

Coolant System [RCS]--Overpressure Protection Systems''; and 3.5.3,

``Emergency Core Cooling Systems--ECCS Subsystems--Tavg  for DOS version 5.1 to

Microsoft Word 97. There were no changes to technical

requirements. The only visible changes to the document are as

follows: (1) the font was changed to Arial 11 point; [(2)] page

numbers were revised to an LCO [limiting condition for operation]

specific numbering scheme; and [(3)] intentionally blank pages were

deleted.

The proposed changes discussed above are strictly

administrative/editorial and do not affect the operation or function

of any plant system, component, or structure. Therefore, the

proposed changes do not increase the probability of occurrence or

the consequences of a previously evaluated accident.

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

and different type of accident from any previously evaluated.

The proposed administrative/editorial changes do not alter the

operation of any plant system or equipment and do not introduce a

new mode of operation. Each requirement contained in the license

conditions proposed for deletion has either been completed or is

obsolete. Since these parts of the license are no longer applicable,

deletion of these items does not provide the potential for an

accident to be created. The conversion of the Technical

Specifications from one word processing format to another did not

involve any changes to technical requirements. Thus, the proposed

changes cannot create a new accident initiating mechanism, and do

not create the possibility of a new and different type of accident

from any previously evaluated.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The license conditions proposed for deletion are obsolete and

each requirement has been completed. The conversion of the Technical

Specifications from one word processing format to another did not

involve any changes to technical requirements. Since the proposed

changes are strictly administrative/editorial and do not involve any

physical or procedural changes to the plant, the margin of safety,

as defined in the bases for any Technical Specification is not

affected by the proposed changes.

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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia.

NRC Project Director: Herbert N. Berkow.

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

Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: November 16, 1996 (TS 98-06).

Brief description of amendments: The proposed amendments would

change the Sequoyah Nuclear Plant Technical Specifications (TSs) by

revising the emergency diesel generator (EDG) surveillance requirements

(SRs) to add a note that allows the SR to be performed in Modes 1, 2, 3

or 4, if the associated components are already out-of-service for

testing or maintenance and to remove the SR that verifies certain

lockout features prevent EDG starting.

Basis for proposed no significant hazards consideration

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

Authority (TVA), the licensee, has provided its analysis of the issue

of no significant hazards consideration, which is presented below:

TVA has concluded that operation of SQN Units 1 and 2, in

accordance with the proposed change to the TSs, does not involve a

significant hazards consideration. TVA's conclusion is based on its

evaluation, in accordance with 10 CFR 50.91(a)(1), of the three

standards set forth in 10 CFR 50.92(c).

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The probability of occurrence or the consequences for an

accident or malfunction of equipment is not increased by this

request. The proposal does not alter the way any structure, system

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

plant is operated, and does not alter equipment out-of-service time.

This request does not degrade the ability of the D/G [emergency

diesel generator] or equipment downstream of the load sequencers to

perform their intended function. Deleting the surveillance of a

nonsafety-related equipment protection function from TS likewise

does not change the probability or consequences of analyzed accident

scenarios. Dose consequences remain unchanged by this request.

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

new or different kind of accident from any accident previously

evaluated.

A possibility for an accident or malfunction of a different type

than any evaluated previously in SQN's FSAR [Final Safety Analysis

Report] is not created; nor is the possibility for an accident or

malfunction of a different type. The proposal does not alter the way

any structure, system or component functions and does not modify the

manner in which the plant is operated.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The margin of safety has not been reduced since the test

methodologies are not being changed and LCO [Limiting Condition for

Operation] allowed outage times are not being changed. Deleting the

surveillance of a nonsafety-related equipment protection function

from TS likewise does not reduce the margin of safety. The results

of accident analysis remain unchanged by this request.

The NRC 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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

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

Plant, Unit 1, Lake County, Ohio

Date of amendment request: October 27, 1998.

Description of amendment request: The proposed amendment would

modify the existing Minimum Critical Power Ratio (MCPR) Safety Limit

contained in Technical Specification 2.1.1.2. The change would apply

additional conservatism by modifying the MCPR Safety Limit values, as

calculated by General Electric, by maintaining the limit of 1.09 for

two recirculation loop operation and by increasing the limit from 1.10

to 1.11 for single loop 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.

There is no change to any plant equipment. Per USAR Section

4.2.1, the fuel system design bases are provided in General Electric

Standard Application for Reactor Fuel (GESTAR II). The Minimum

Critical Power Ratio (MCPR) Safety Limit protects the fuel in

accordance with the design basis. The MCPR Safety Limit calculations

limit the bundle power to ensure the critical power ratio remains

unchanged. Therefore, there is not an increase in the probability of

transition boiling. The basis of the MCPR Safety Limit calculation

remains the same,

[[Page 66604]]

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

transition boiling if the limit is not violated. Therefore, there is

no increase in the probability of the occurrence of a previously

analyzed accident.

The fundamental sequences of accidents and transients have not

been altered. The MCPR Operating Limits are selected such that

potentially limiting plant transients and accidents prevent the MCPR

from decreasing below the MCPR Safety Limit anytime during the

transient. Therefore, there is no impact on any of the limiting USAR

Appendix 15B transients. The radiological consequences are the same

as previously stated in the USAR, and as approved in the NRC Safety

Evaluation for GESTAR II. Therefore, the consequences of an accident

do not increase over previous evaluations in the USAR.

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

or different kind of accident from any accident previously

evaluated.

The MCPR Safety Limit values are designed to ensure that fuel

damage from transition boiling does not occur in at least 99.9% of

the fuel rods in the core as a result of the limiting postulated

accident. The values are calculated in accordance with GESTAR II and

the fuel vendor's interim implementing procedures, which incorporate

cycle-specific parameters.

The GESTAR II analysis has been accepted by the NRC as

comprehensive for ensuring that fuel designs will perform within

acceptable bounds. The MCPR Safety Limit ensures that the fuel is

protected in accordance with the design basis. The function,

location, operation, and handling of the fuel remain unchanged. In

addition, the initiating sequence of events has not changed.

Therefore, no new or different kind of accident is created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The MCPR Safety Limit values do not alter the design or function

of any plant system, including the fuel. The new MCPR Safety Limit

values were calculated using NRC-approved methods described in

GESTAR II and the fuel vendor's interim implementing procedures,

which incorporate cycle-specific parameters. The MCPR Safety Limit

values are consistent with GESTAR II, the NRC Safety Evaluation of

GESTAR II, the NRC Safety Evaluation Report for the Perry Nuclear

Power Plant and its Supplements for USAR Sections 4.4.1 and

15.0.3.3.1, and the Technical Specification Bases (Section 2.1.1.2)

for the MCPR Safety Limit. This change incorporates a cycle-specific

MCPR Safety Limit, as opposed to relying on the generic limit.

Therefore, the implementation of the proposed change to the MCPR

Safety Limit does not involve a reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

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

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

NRC Project Director: Stuart A. Richards.

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

Callaway County, Missouri

Date of application request: October 27, 1998 (supersedes the April

12, 1996, amendment request). This notice supersedes the staff's

proposed no significant hazards consideration determination evaluation

for the requested changes that was published on May 8, 1996 (61 FR

20858).

Description of amendment request: The proposed amendment

application would change the technical specifications (TS) for the

reactor coolant system and associated Bases to allow the installation

of electrosleeves in the Callaway steam generators for two fuel cycles.

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 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 sleeve are bounded by the limits established in the ASME

Code. 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

RG [Regulatory Guide] 1.121. Leakage testing for \5/8\'', \7/8\'',

\11/16\'' and \3/4\'' tube sleeves has demonstrated that no

unacceptable levels of primary to secondary leakage are expected

during any plant condition.

The sleeve nominal wall thickness (used for developing the

depth-based plugging limit for the sleeve) 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, which applies to part throughwall

degradation, is that the minimum acceptable wall must maintain a

factor of safety of three against tube failure under normal

operating (design) 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 condition requirement minimum

wall thickness.

A bounding tube wall degradation growth rate per cycle and a NDE

[Non-Destructive Examination] uncertainty has been assumed for

determining the sleeve TS plugging limit. The sleeve wall

degradation extent is determined by NDE. The degradation which would

require plugging sleeved tubes is developed using the guidance of RG

1.121 and is defined in BAW-10219P, to be 20% throughwall for any

service induced degradation.

The consequences of failure of the sleeve are bounded by the

current steam generator tube rupture analysis included in the

Callaway FSAR [Final Safety Analysis Report]. Due to the slight

reduction in diameter caused by the sleeve wall thickness, primary

coolant release rates would be slightly less than assumed for the

steam generator tube rupture analysis (depending on the break

location), and therefore, would result in lower total primary fluid

mass release to the secondary system.

A risk assessment for installation of Electrosleeves at Callaway

Plant was performed for a two-cycle operating period. The results of

this evaluation determined that sufficient margins against

postulated tube rupture during bounding accident conditions exist

for all types of degradation of the Electrosleeve material. The

calculated probability of burst for a hypothetical population of

10,000 axial flaws, 100% throughwall of the parent tube and 0.40''

long, is 4.4 x 10-11 at the end of the second operating cycle. The

probability of burst for postulated circumferential flaws and pits

is determined to be essentially zero.

The proposed change does not adversely impact any other

previously evaluated design basis accident or the results of LOCA

[Loss of Coolant Accident] and non-LOCA accident analyses for the

current technical specification minimum reactor coolant system flow

rate. The results of the analyses and testing demonstrate that the

electrosleeve is an acceptable means of maintaining tube integrity.

Furthermore, per Regulatory Guide 1.83 recommendations, the sleeved

tube can be monitored through periodic inspections with present NDE

techniques. These measures demonstrate that installation of sleeves

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

condition consistent with its original design basis.

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

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

[[Page 66605]]

Electrosleeving does not represent a potential to 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 to that of the originally supplied tubing during

all plant conditions. Leak and mechanical testing of electrosleeves

support the conclusions of the calculations that each sleeve retains

both structural and leakage integrity during all conditions.

Sleeving of tubes does not provide a mechanism resulting in an

accident outside of the area affected by the sleeves. Any accident

as a result of potential tube or sleeve degradation in the repaired

portion of the tube is bounded by the existing tube rupture accident

analysis.

Implementation of sleeving will reduce the potential for primary

to secondary leakage during a postulated steam line break 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 steam line break leakage is

reduced. These degraded intersections now are returned to a

condition consistent with the Design Basis. While the installation

of a sleeve reduces primary coolant flow, the reduction is far below

that caused by plugging. Therefore, far greater primary coolant flow

area is maintained through sleeving versus plugging.

3. The proposed change does not 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, i.e., tube/

sleeve operational and faulted condition stresses are bounded by the

ASME Code requirements and the repaired tubes are leaktight. The

safety factors used in the design of sleeves 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

sleeve assembly which represent the reactor coolant pressure

boundary can be monitored for the initiation and progression of

sleeve/tube wall degradation, thus satisfying the requirements of

Regulatory Guide 1.83. The portion of the tube bridged by the sleeve

is effectively removed from the pressure boundary, and the sleeve

then forms the new pressure boundary. The areas of the sleeved tube

assembly which require inspection are defined in BAW-10219P.

In addition, since the installed sleeve represents a portion of

the pressure boundary, a baseline inspection of these areas is

required prior to operation with sleeves installed. The effect of

sleeving on the design transients and accident analyses has been

reviewed based on the installation of sleeves up to the level of

steam generator tube plugging coincident with the minimum reactor

flow rate and the Callaway Safety Analysis.

Provisional requirements cited in other NRC Safety Evaluation

Reports addressing the implementation of sleeving have required the

reduction of the individual steam generator normal operation primary

to secondary leakage limit from 500 to 150 gpd [gallons per day].

Consistent with these evaluations, Union Electric will reduce the

per steam generator leak rate of 500 gpd in TS 3.4.6.2.c to 150 gpd.

The establishment of this leakage limit at 150 gpd provides

additional safety margin. [The staff notes that this leakage limit

has been incorporated into the Callaway Technical Specifications via

license amendment #119 dated October 1, 1996.]

Finally, Union Electric will reduce the tube plugging limit from

48% through wall to 40% through wall to be consistent with NUREG-

1431. The establishment of the plugging limit at 40% through wall

provides additional safety margin. [The staff notes that this

plugging limit has been incorporated into the Callaway Technical

Specifications via license amendment #119 dated October 1, 1996.]

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: University of Missouri-

Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

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

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

NRC Project Director: William H. Bateman.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: November 3, 1998.

Description of amendment request: The licensee proposes to make

administrative changes to the Technical Specifications to correct

errors, add consistency within the Technical Specifications, and make

nomenclature changes to support and enhance usability of the Technical

Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, because:

The proposed changes are purely administrative in nature and

have no effect on plant hardware, plant design, safety limit

setting, or plant system operation and therefore do not modify or

add any initiating parameters that would significantly increase the

probability or consequences of an accident previously evaluated.

No new modes of operation are introduced by the proposed changes

such that adverse consequences would result. Accordingly, the

consequences of previously analyzed accidents are not affected by

this proposed license amendment.

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

from any accident previously evaluated, because:

These changes do not affect the operation of any systems or

components, nor do they involve any potential initiating events that

would create any new or different kind of accident. Therefore, the

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

kind of accident from any accident previously evaluated for the

Vermont Yankee Nuclear Power Station.

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

because:

These proposed changes do not affect any equipment involved in

potential initiating events or safety limits. Therefore, it is

concluded that the proposed changes do not involve a significant

reduction in a margin of safety.

Administrative changes, as such, do not constitute any

significant hazards considerations.

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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Project Director: Cecil O. Thomas.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: November 10, 1998.

Description of amendment request: The proposed changes to North

Anna Power Station (NAPS), Units 1 and 2, Technical Specification (TS)

3.4.4 will clarify the operability requirements for the pressurizer

heaters and eliminate a potential verbatim compliance issue associated

with the pressurizer heaters and emergency power supply. The verbatim

compliance issue was created when the Emergency Diesel Generator

allowed outage time was changed from 72 hours to 14 days.

Basis for proposed no significant hazards consideration

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

[[Page 66606]]

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

Virginia Electric and Power Company has reviewed the

requirements of 10 CFR 50.92 as they relate to the proposed changes

for the North Anna Units 1 and 2 and determined that a significant

hazards consideration is not involved. The proposed changes will

revise the LCO [limiting condition for operation] 3.4.4 to require

that the pressurizer have two groups of pressurizer heaters operable

with a capacity of greater than or equal to 125 kW and capable of

being powered from its associated emergency bus. The Action

Statement will also be revised to focus on heater operability. The

following is provided to support this conclusion.

(a) Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The pressurizer heaters are not an initiator of any accident

previously evaluated. As a result, the probability of any accident

previously evaluated is not increased. The pressurizer heaters

remain operable as assumed in the accident analysis to mitigate the

consequences of any accident. Therefore, the proposed changes to

clarify the operability requirements do not significantly increase

the probability of occurrence or the consequences of any previously

analyzed accident.

(b) Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed Technical Specifications changes do not involve any

physical alteration of the plant or changes in methods governing

normal plant operation. Operation of and the design of the

pressurizer heaters and the associated power supplies are not

changed by the proposed changes. The proposed changes do not impose

any new or eliminate any existing requirements. Therefore, it is

concluded that no new or different kind of accident or malfunction

from any previously evaluated has been created.

(c) Does the change involve a significant reduction in a margin

of safety?

The proposed Technical Specifications changes will not reduce

the margin of safety since the change has no effect on any safety

analyses assumptions. The pressurizer heaters remain operable as

assumed in the safety analysis to mitigate the consequences of any

accident previously analyzed. The proposed changes only clarify the

operability requirements for the pressurizer heaters and associated

emergency power supplies. Therefore, the proposed changes do not

result in a significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,

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

23219.

NRC Project Director: Herbert N. Berkow.

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.

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

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

Date of amendment request: October 29, 1998.

Description of amendment request: Technical Specification changes

(TS) relating to the implementation and automatic removal of certain

reactor protection system trip bypasses to ensure that the meaning of

explicit terms used in the TSs are consistent with the intent of the

stated requirements.

Date of publication of individual notice in the Federal Register:

November 5, 1998 (63 FR 59809).

Expiration date of individual notice: November 19, 1998.

Local Public Document Room location: Indian River Junior College

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

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

Date of application for amendment: July 20, 1998.

Brief description of amendment: The amendment implements a

modification that constitutes an unreviewed safety question as

described in 10 CFR 50.59. The modification involves replacing the

service water heat exchangers with new plate and frame heat exchangers

having an increased thermal performance capability. The planned

modification is similar to the one completed on Unit 1. In addition, by

a separate letter dated July 20, 1998, the licensee submitted a request

to obtain approval for a temporary one time cooling lineup needed to

support emergency diesel generator operability for the installation of

the Unit 2 service water heat exchanger replacement, which is currently

being reviewed by the NRC

[[Page 66607]]

staff. Therefore, since the implementation of the proposed service

water heat exchanger modification is dependent on the staff's issuance

of the one time Technical Specification (TS) change regarding

installation of the modification, this modification should not be

implemented prior to the issuance of the one-time TS change for

installing the modification.

Date of issuance: November 5, 1998.

Effective date: This license amendment is effective as of the date

of its issuance to be implemented after the staff's issuance of the

one-time TS change regarding the installation of the service water heat

exchanger modification.

Amendment No.: 203.

Facility Operating License No. DPR-69: Amendment revised the

Updated Final Safety Analysis Report.

Date of initial notice in Federal Register: August 12, 1998 (63 FR

43201).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 5, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

Station, Plymouth County, Massachusetts

Date of application for amendment: June 26, 1998.

Brief description of amendment: The amendment modifies various

Technical Specification pages to correct typographical errors, remove

inadvertent replication of information, and updates various Bases

sections.

Date of issuance: November 10, 1998.

Effective date: November 10, 1998.

Amendment No: 178.

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

Technical Specifications.

Date of initial notice in Federal Register: September 23, 1998 (63

FR 50933).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 10, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of application for amendment: March 6, 1998, as supplemented

September 11, 1998. The September 11, 1998, supplemental letter

contained clarifying information only, and did not change the no

significant hazards consideration determination.

Brief description of amendment: The amendment revises Technical

Specification 3.9.2 relating to the use of Post-Accident Monitoring

Source Range neutron flux detectors as a compensatory measure in the

event that one of the two required BF3 neutron flux detectors becomes

inoperable during Mode 6 operations (refueling).

Date of issuance: November 12, 1998.

Effective date: November 12, 1998.

Amendment No: 180.

Facility Operating License No. DPR-23: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 3, 1998 (63 FR

30262).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 12, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

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

Date of application of amendments: September 17, 1998, as

supplemented October 15, 1998.

Brief description of amendments: The amendments revised the Updated

Final Safety Analysis Report to perform a Keowee Emergency Power

Engineered Safeguards Functional Test during the 1998 Unit 3 refueling

outage at Oconee.

Date of Issuance: November 12, 1998.

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

the 1998 Unit 3 refueling outage.

Amendment Nos.: Unit 1--233; Unit 2--233; Unit 3--232.

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

Amendments revised the Updated Final Safety Analysis Report.

Date of initial notice in Federal Register: September 30, 1998 (63

FR 52304).

The October 15, 1998, letter provided clarifying information that

did not change the scope of the September 17, 1998, 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 November 12, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

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

Station, Dauphin County, Pennsylvania

Date of application for amendment: December 2, 1996.

Brief description of amendment: This amendment would revise audit

frequency requirements and relocate them from the Technical

Specifications to the Quality Assurance Plan.

Date of issuance: November 12, 1998.

Effective date: This amendment is effective immediately to be

implemented written 60 days.

Amendment No.: 52.

Facility Operating License No. DPR-73: The amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: July 30, 1997 (62 FR

40850).

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San

Diego County, California

Date of application for amendments: May 11, 1998, as supplemented

by letter dated October 9, 1998.

Brief description of amendments: The amendments modify the

technical specifications (TS) for San Onofre Nuclear Generating Station

Unit Nos. 2 and 3 to implement 10 CFR Part 50 Appendix J, Option B for

performance-based reactor containment leakage testing.

Date of issuance: November 6, 1998.

Effective date: November 6, 1998, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 2 -144; Unit 3 -135.

Facility Operating License No. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

[[Page 66608]]

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48265).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 6, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

California, P. O. Box 19557, Irvine, California 92713.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: July 6, 1998.

Brief description of amendments: Relocates the description of the

reactor coolant system design features in Technical Specification 5.4

to the Updated Final Safety Analysis Report, which already contains the

information.

Date of issuance: November 18, 1998.

Effective date: November 18, 1998, to be implemented within 30

days.

Amendment Nos.: Unit 1--Amendment No. 98; Unit 2--Amendment No. 85.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48266).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 18, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas.

Date of amendment request: July 6, 1998 , as supplemented on

October 28, 1998.

Brief description of amendments: Relocate the Technical

Specification 3/4.3.3.3 requirements for Seismic Instrumentation to the

Technical Requirements Manual.

Date of issuance: November 18, 1998.

Effective date: November 18, 1998, to be implemented within 30

days.

Amendment Nos.: Unit 1--Amendment No. 99; Unit 2--Amendment No. 86.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48267).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 18, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas.

Date of amendment request: July 6, 1998, as supplemented on October

28, 1998.

Brief description of amendments: Relocates the Technical

Specification 3/4.7.13 requirements for the Area Temperature Monitoring

System to the Technical Requirements Manual.

Date of issuance: November 18, 1998.

Effective date: November 18, 1998, to be implemented within 30

days.

Amendment Nos.: Unit 1--Amendment No. 100; Unit 2--Amendment No.

87.

Facility Operating License Nos. NPF-76 and NPF-80: The amendment

revises the Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48267). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 18, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

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: February 13, 1998 (TS 97-07).

Brief description of amendments: The amendments incorporate new

main steam isolation valve (MSIV) requirements that are consistent with

NUREG-1431, the Westinghouse Standard Technical Specifications (TS),

including testing requirements for the MSIVs that ensure the valves

close on an automatic actuation signal.

Date of issuance: November 17, 1998.

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

later than 45 days after issuance.

Amendment Nos.: 236 and 226.

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

revise the technical specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19980).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 17, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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: June 26, 1998 (TS 98-02).

Brief description of amendments: The amendments change the

Technical Specifications and their Bases to lower the specific activity

limit for the primary coolant system from 1.0 microcurie/gram dose

equivalent iodine-131 to 0.35 microcurie/gram, as provided for in NRC

Generic Letter 95-05, ``Voltage-Based Repair Criteria for Westinghouse

Steam Generator Tubes Affected by Outside Diameter Stress Corrosion

Cracking.'' This change allows a proportional increase in main steam

line break induced primary-to-secondary leakage when implementing the

alternate steam generator tube repair criteria, which the NRC has

already approved for Sequoyah Nuclear Plant, Units 1 and 2.

Date of issuance: November 17, 1998.

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

later than 45 days after issuance.

Amendment Nos.: 237 and 227.

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

revise the technical specifications.

Date of initial notice in Federal Register: July 15, 1998 (63 FR

38205).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 17, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

[[Page 66609]]

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,

Unit 1, (WBN) Rhea County, Tennessee

Date of application for amendment: August 5, 1998 (TS 98-008).

Brief description of amendment: This amendment is in response to

your application dated August 5, 1998. The amendment revises the WBN

Technical Specifications (TS) and associated TS Bases to allow up to 4

hours to make the residual heat removal suction relief valve available

as a cold overpressure mitigation system relief path.

Date of issuance: November 10, 1998.

Effective date: November 10, 1998.

Amendment No.: 14.

Facility Operating License No. NPF-90: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: September 23, 1998 (63

FR 50940).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 10, 1998.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: July 10, 1996 (TXX-96405), as

supplemented by letters dated October 1, 1996 (TXX-96475), and July 1,

1998 (TXX-98159).

Brief description of amendments: The amendment would take credit

for the addition of train oriented Fan Coil Units for each UPS and

Distribution Room and would provide redundancy to the existing Air

Conditioning (A/C) Units (TS 3/4.7.11 and its associated bases).

Date of Issuance: Date of issuance: November 18, 1998.

Effective date: November 18, 1998, to be implemented within 30

days.

Amendment Nos.: Unit 1--Amendment No. 61; Unit 2--Amendment No. 47.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

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

FR 6579).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 18, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019.

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: May 7, 1998.

Brief description of amendment: This amendment revises Technical

Specification 5.4, ``Fuel Storage,'' to increase the allowable mass of

uranium-235 (U235) per axial centimeter for fuel storage.

The requested change will allow the use of new Siemens Power

Corporation heavy fuel assembly designs.

Date of Issuance: November 12, 1998.

Effective date: November 12, 1998.

Amendment No.: 141.

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

Technical Specifications.

Date of initial notice in Federal Register: June 17, 1998 (63 FR

33111).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 12, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

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

Dated at Rockville, Maryland, this 24th day of November 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

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

Reactor Regulation.

[FR Doc. 98-31931 Filed 12-1-98; 8:45 am]

BILLING CODE 7590-01-P