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) 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, 1999, through November 19, 1999.

The last biweekly notice was published on November 17, 1999 (64 FR

62704).

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

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

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By January 3, 2000, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and

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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 electronically from the ADAMS Public Library

component on the NRC Web site, http://www.nrc.gov (the Electronic

Reading Room). If a request for a hearing or petition for leave to

intervene is filed by the above date, the Commission or an Atomic

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

Chairman of the Atomic Safety and Licensing Board Panel, will rule on

the request and/or petition; and the Secretary or the designated Atomic

Safety and Licensing Board will issue a notice of a hearing or an

appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

inform the Commission by a toll-free telephone call to Western Union at

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

operator should be given Datagram Identification Number N1023 and the

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

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

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

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 electronically from the ADAMS Public

Library component on the NRC Web site, http://www.nrc.gov (the

Electronic Reading Room).

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

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

Date of amendment request: October 12, 1999.

Description of amendment request: This proposed technical

specification change removes the anticipatory reactor scram signal for

turbine electro-hydraulic control (EHC) low oil pressure trip from the

reactor protection system (RPS) trip function.

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:

Does the change involve a significant increase in the probability

of occurrence or consequences of an accident previously evaluated?

The proposed change removes the Turbine EHC Control Oil

Pressure-Low scram function and the associated Limiting Safety

System Setting (LSSS). The purpose of the Turbine EHC Control Oil

Pressure scram is to anticipate the pressure transient which would

be caused by imminent control valve closure on loss of control oil

pressure. This

[[Page 67332]]

function does not serve as an initiator for any accidents evaluated

in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR).

In addition, this trip function is not credited in any design basis

event and is functionally redundant to the Turbine Control Valve

Fast Closure RPS trip function during a postulated loss of EHC

control oil event. The Turbine Control Valve Fast Closure will

initiate a scram on a loss of control oil event coincident with

turbine control valve closure.

Therefore, this proposed amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The removal of this function does not represent a change in

operating parameters or introduce a new mode of operation. The

pressure switches associated with the Turbine Control Valve Fast

Closure function provide equivalent protection from a loss of EHC

oil event. For this reason, the change does not create the

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

previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

Operation under the proposed amendment will not change any plant

operation parameters, nor any protective system actuation setpoints

other than removal of the Turbine EHC Control Oil Pressure-Low scram

function. The scram function associated with the Turbine Control

Valve Fast Closure provides equivalent protection for events

involving fast turbine control valve closure including the loss of

EHC control oil pressure. For this reason, eliminating the EHC

Control Oil Pressure-Low scram function, which is redundant to other

protective instrumentation, does not reduce the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

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

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

York

Date of amendment request: September 23, 1999.

Description of amendment request: The proposed amendment would

relocate items associated with instrumentation for toxic gas monitoring

from the Technical Specifications (TSs) to the Updated Final Safety

Analysis Report (UFSAR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve a significant hazards

consideration because:

1. There is no significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes are administrative in nature. The

Specifications and associated Bases will be transferred verbatim to

the UFSAR.

These changes do not affect possible initiating events for

accidents previously evaluated or alter the configuration or

operating of the facility. The Limiting Safety Systems Settings and

Safety Limits specified in the current TSs remain unchanged.

Therefore, the proposed changes to the subject TS would not increase

the probability or consequences of an accident previously evaluated.

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

any accident previously evaluated has not been created.

As stated above, the proposed changes are administrative in

nature. The safety analysis of the facility remains complete and

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

plant conditions for which the design basis accidents have been

evaluated are still valid. The operating procedures and emergency

procedures are unaffected. Consequently, no new failure modes are

introduced as a result of the proposed changes, therefore, the

proposed changes will not initiate any new or different kind of

accident.

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

safety.

The proposed changes are administrative in nature. Since there

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

the UFSAR design basis, accident assumptions are not affected.

Therefore, the proposed changes will not result in a reduction in

the margin of safety.

The proposed changes have been reviewed by both the Station

Nuclear Safety Committee (SNSC) and the Con Edison Nuclear Facility

Safety Committee (NFSC). Both Committees concur that the proposed

changes do not represent 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.

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

New York, New York 10003.

NRC Section Chief: Sheri Peterson.

Duke Energy Corporation, 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 3, 1999.

Description of amendment request: The amendments would revise

Section 3.8.1, ``AC [alternating current] Sources--Operating,'' of the

Technical Specifications. Specifically, this would revise: (1)

Surveillance Requirement (SR) 3.8.1.9 to delete the power factor

requirement from the diesel generator (DG) load rejection test; (2) SR

3.8.1.13 to allow performance of the diesel generator non-emergency

automatic trip bypass test at any operational power level; and (3) SR

3.8.1.14 to allow performance of the 24-hour diesel generator run at

any operational power level and delete the power factor requirement. No

plant modification is involved with this proposed amendment.

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, or

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

from any accident previously evaluated, or

3. Involve a reduction in a margin of safety.

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 effect

on accident probabilities or consequences. The DGs and their

associated emergency buses are not accident initiating equipment;

therefore, there will be no impact on any accident probabilities by

the approval of this amendment. The design of the equipment is not

being modified by these proposed changes. In addition, the ability

of the DGs to respond to a design basis accident will not be

adversely impacted by these proposed changes. There will be no

significant increased likelihood of causing a blackout of a safety

bus by the proposed changes in testing. Therefore, there will be no

significant 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. Equipment

will be operated in the same configuration with the exception of the

plant

[[Page 67333]]

mode in which the testing is conducted. 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; neither does it adversely impact any accident

mitigating systems.

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 equipment referenced in the revised TS for

these proposed changes 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.

Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department

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

North Carolina 28201-1006.

NRC Section Chief: Richard L. Emch, Jr.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear

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

Date of amendment request: November 3, 1999.

Description of amendment request: The proposed amendments would

revise Section 3.8.1, ``AC [alternating current] Sources--Operating,''

of the Technical Specifications. Specifically, this would revise: (1)

Surveillance Requirement (SR) 3.8.1.9 to allow performance of the

diesel generator (DG) load rejection test at any operational power

level and to delete the power factor requirement; (2) SR 3.8.1.10 to

allow performance of the diesel generator full load rejection test at

any operational power level; and (3) SR 3.8.1.14 to allow performance

of the 24-hour diesel generator run at any operational power level and

delete the power factor requirement. No plant modification is involved

with this proposed amendment.

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, 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.

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 effect

on accident probabilities or consequences. The DGs and their

associated emergency buses are not accident initiating equipment;

therefore, there will be no impact on any accident probabilities by

the approval of this amendment. The design of the equipment is not

being modified by these proposed changes. In addition, the ability

of the DGs to respond to a design basis accident will not be

adversely impacted by these proposed changes. There will be no

significant increased likelihood of causing a blackout of a safety

bus by the proposed changes in testing. Therefore, there will be no

significant 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. Equipment

will be operated in the same configuration with the exception of the

plant mode in which the testing is conducted. 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; neither does it adversely impact any

accident mitigating systems.

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 equipment referenced in the revised TS for

these proposed changes 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.

Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department

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

North Carolina 28201-1006.

NRC Section Chief: Richard L. Emch, Jr.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of amendment request: October 7, 1999.

Description of amendment request: Grand Gulf Nuclear Station (GGNS)

requests approval to revise its licensing basis for the release of

fission products following an accident. The basis for the proposed

change makes use of one of the insights established in NUREG-1465,

``Accident Source Terms for Light Water Nuclear Power Plants,'' which

defines alternative source terms for use in the licensing of light

water reactors. Specifically, this application credits the insight that

there is a delay in the release of fission products from the reactor

fuel following a postulated design basis loss-of-coolant accident

(LOCA). The timing of fission product release from fuel perforation,

i.e., gap activity release, is based on the boiling water reactor

(BWR)--specific value of the timing of the gap activity release phase

of a LOCA as calculated in the Boiling Water Reactor Owners Group

(BWROG) Report, ``Prediction of the Onset of Fission Gas Release From

Fuel in Generic BWR.'' This BWROG Report has been previously reviewed

and approved by the Nuclear Regulatory Commission (NRC) staff. The

licensing basis change to Updated Final Safety Analysis Report (UFSAR)

Section 15.6.5.5.2 proposed by GGNS replaces the assumption of an

instantaneous release of gap activity phase fission products into the

drywell with a more accurate scenario in which the gap activity release

is delayed by up to 121 seconds as calculated in the BWROG Report.

Approval of this change will allow GGNS to increase the containment

isolation valve closure times credited for limiting post-accident doses

to both control room personnel and to offsite individuals. While this

new basis would be applicable to all of the containment isolation

valves, it addresses only the dose mitigation aspects of the closure

requirements. There are currently some valves for which the closure

time is limited based on other functional performance requirements

(e.g., line break isolation). This submittal does not propose any

changes that would

[[Page 67334]]

eliminate any of these other requirements. The allowable closure times

for these valves would not be affected by this proposed change.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

GGNS staff has evaluated the proposed change to incorporate a

delay in the post-accident fission product release into its

licensing basis. This change recognizes one of the revised source

term insights discussed in NUREG-1465. This change in the licensing

basis will provide the basis for revising the Technical Requirements

Manual to increase Primary Containment Isolation Valve (PCIV)

maximum isolation times. These changes have been evaluated using the

standards in 10CFR50.92 and it is concluded that they do not involve

any significant hazards considerations. Specifically, the proposed

change will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated,

The proposed change takes credit for a new source term insight

that recognizes that the fission product release from a fuel

assembly is not instantaneous with a design basis accident.

Implementation of this change into the licensing basis will be used

to justify an increase in the maximum allowable PCIV isolation

times. These changes do not affect the precursors for any accident

or transient evaluated in Chapter 15 of the GGNS UFSAR. Therefore,

there is no increase in the probability of any accident previously

evaluated.

A plant specific radiological analysis has been performed to

evaluate the effect on the dose consequences of extending the

maximum allowable closure time. This evaluation considered the

initial two-minute period of the accident during which, according to

new source term insights developed in NUREG-1465 and in a BWROG

report, fission product releases are not expected to occur. Releases

from the break and from containment during this period consist of

coolant radioactivity only. The total release during this period was

found to result in an offsite dose of less than 0.60 rem. This dose

represents only a small fraction of the LOCA dose evaluated in the

UFSAR. As this submittal is for a limited scope application of the

NUREG-1465 insights (in this case, timing and duration of the

coolant activity phase) and addresses only the first 121 seconds of

the accident scenario, the total long-term dose determined using the

TID-14844 assumptions is not changed by this submittal.

In reality, the other insights offered in the NUREG would be

expected to result in an overall dose reduction. In any event, the

dose consequences of the proposed change do not result in an

increase in the consequences of any accident previously evaluated.

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

accident from any accident previously evaluated;

The primary containment isolation system is designed to prevent,

as much as practicable, the unfiltered release of radioactive

material to the environs following an accident. As such, the system

is relied upon for accident dose consequence mitigation. Neither the

revision of the licensing basis to recognize that fission product

releases are not instantaneous as is assumed in the current

analysis, nor the extension of the valve closure times affects the

ability of the valves to perform their accident mitigation function.

It is also noted that the increased closure time allowables will

only be applied to valves which do not have an alternate

constraining performance requirement for closure time; the safety

functions of other supported components and systems are not

affected. Thus, the proposed change does not create the potential

for a new or different kind of accident.

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

The proposed change revises the bases for the offsite dose

calculation to credit, in the initial 2 minutes of the accident

scenario, the fact that there is no fuel failure expected during

this time. That is, for the first two minutes of the event, only

coolant activity is released. The other assumptions, bases and

methodologies for offsite dose calculations used to evaluate the

long-term offsite dose consequences of accidents described in FSAR

[Final Safety Analysis Report] Chapter 15 are not affected by this

change. The margin between calculated dose consequences described in

the FSAR and regulatory limits is not reduced.

A recent GGNS analysis of the LOCA scenario considering the only

release in the first 121 seconds is from the reactor coolant

resulted in an EAB [exclusion area boundary] dose of less than 1 rem

thyroid during this period. The total dose for the 0- to 2-hour

period is not expected to increase due to the delay in the fission

product release; the total amount of radioactivity released will

remain the same. Both the recently evaluated 2-minute dose and the

24.9 rem in two hours as presented in the UFSAR are insignificant in

comparison to the 300 rem acceptance limit for this scenario. The

GGNS SER [safety evaluation report] acknowledges the conservatism of

the old analysis methodology. An independent analysis done by the

staff during their evaluation of the GGNS FSAR estimated doses could

decrease about 95% if the fission product release were to be delayed

by 2 minutes.

The bases for PCIV closure times described in the Technical

Specifications remain unchanged. The inconsistency between the

assumption of immediate containment isolation in the dose analysis

and allowable isolation valve closure times of one to two minutes is

eliminated by this change. Plant specific analysis has shown that

the expected dose resulting from the PCIVs remaining open during

this period is insignificant.

Actual safety benefits are expected to result from valve

performance and reliability improvements, elimination of unnecessary

reports and system performance improvements such as minimization of

water hammer events. Therefore, the increase in maximum isolation

time for certain PCIVs proposed in this submittal will not result in

a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.

NRC Section Chief: Robert A. Gramm.

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

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

Date of amendment request: August 20, 1999.

Description of amendment request: The proposed license amendment

would modify the Technical Specifications (TSs) to allow revision of

the 4KV Engineered Safeguards Bus Undervoltage Relay Degraded Voltage

calibration to be performed at an annual interval rather than its

present refueling interval and change the bases to state that the

degraded voltage relay setpoint tolerance is being changed from an ``as

left'' reading to an ``as found'' reading. Additionally, the new

calculations supporting the request identified a need to compensate for

lack of voltage margin through reliance on manual action in lieu of

full automatic voltage protection, as implied by Chapter 8 of the

Updated Final Safety Analysis Report (UFSAR). Such actions would

involve load manipulations following a loss of coolant accident (LOCA)

with post LOCA conditions in combination with extremely low switchyard

voltage. An additional limit of operation with a maximum of 5

Circulating Water pumps while in single 230KV auxiliary transformer

operation is also added to the UFSAR.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes to the degraded voltage relay setpoint

tolerance and calibration interval are intended to reduce the total

degraded voltage relay setpoint uncertainties. These changes will

provide greater confidence that minimum voltages necessary to

operate NSR [nuclear safety related] equipment are not exceeded. In

combination, the proposed changes for degraded voltage relay

setpoint tolerance and

[[Page 67335]]

calibration interval will reduce the probability that ES [engineered

safeguards] buses will be separated from their offsite power source

during low grid voltage conditions. This will reduce challenges to

the onsite emergency power systems. The proposed changes will

enhance the ability of the undervoltage protection scheme to perform

in accordance with its intended design, and will improve the ability

of the scheme to respond to low voltage conditions caused by

malfunction of equipment important to safety.

Therefore, operation of the facility in accordance with the

proposed amendment will not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated in the SAR.

2. The proposed setpoint tolerance and calibration interval

changes are consistent with the specifications and intended design

of the degraded voltage protection scheme and do not introduce the

possibility of any new failure modes to the protection scheme or the

electrical distribution system. The proposed changes reduce the

probability of insufficient voltage to NSR loads and reduce the

probability of separation of ES buses from the offsite power source.

Therefore, operation of the facility in accordance with the proposed

changes do not create a possibility of a new or different type of

accident than any previously evaluated in the SAR.

3. The proposed setpoint tolerance and calibration interval

changes are intended to reduce the total degraded voltage relay

setpoint uncertainties. The changes will provide greater confidence

that minimum voltages necessary to operate NSR equipment will not be

exceeded. The proposed changes will also reduce the probability that

the ES buses will be separated from their offsite power source

during low grid voltage conditions. These effects will enhance the

objective [of] providing a reliable source of power for BOP

auxiliaries and [a] continuously available power supply for the ES

equipment as required by TS [technical specification] 3.7 bases.

Therefore, operation of the facility in accordance with the proposed

changes would not involve a significant reduction in a margin of

safety.

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

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.

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

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

NRC Section Chief: Sheri R. Peterson.

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

C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment requests: November 3, 1999.

Description of amendment requests: The proposed amendments would

allow use of fuel rods with ZIRLO cladding, specify an alternate

methodology to determine the integral fuel burnable absorber (IFBA)

requirements for Westinghouse fuel assemblies stored in the new fuel

storage racks, and delete the designation of the fuel assembly types

allowed in the spent fuel storage racks and the new fuel storage racks.

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 of occurrence or consequences of an accident previously

evaluated?

The proposed T/S [Technical Specification] change to allow

storage and use of fuel rods clad with ZIRLO does not significantly

increase the probability of occurrence of an accident. Fuel

assemblies are not an initiator or precursor to any previously

evaluated accident. The proposed T/S change does not change or alter

the design criteria for the systems or components used to mitigate

the consequences of any design basis accident. Use of ZIRLO fuel

cladding does not adversely affect fuel performance or impact

nuclear design methodology. Therefore, accident analysis results are

not impacted. The operating limits are not changed and the analysis

methods to demonstrate operation within the limits remain in

accordance with NRC-approved methodologies. Other than the changes

to the fuel rod cladding there are no physical changes to the plant

associated with this T/S change. A safety analysis is still required

to be performed for each specific reload cycle to demonstrate

compliance with fuel safety design bases. The 10 CFR 50.46 emergency

core cooling system acceptance criteria are applied to the ZIRLO

clad fuel rods. The use of fuel assemblies containing ZIRLO clad

fuel rods does not result in a change to the reload design and

safety analysis limits. The clad material is similar in chemical

composition and has similar physical and mechanical properties as

Zircaloy-4. Thus, the cladding integrity is maintained and the

structural integrity of the fuel assembly is not affected. ZIRLO

cladding improves corrosion performance and dimensional stability.

Since the dose predictions in the safety analyses are not sensitive

to the fuel rod cladding material used, the radiological

consequences of accidents previously evaluated in the safety

analysis remain valid.

The proposed T/S change to specify an alternate NRC-approved

methodology used to determine the IFBA requirements for Westinghouse

fuel assemblies stored in the new fuel storage racks does not change

or alter the design criteria for the systems or components used to

mitigate the consequences of any design basis accident. This

alternate methodology is more conservative with respect to

determining the reactivity of the stored fuel assemblies than the

methodology currently specified in the T/S. Therefore, the

probability of an accidental criticality is less with the proposed

T/S change than currently assumed. Since a criticality accident is

precluded by the proposed T/S change, the consequences of a

criticality accident are not changed by the use of this alternate

methodology.

The proposed T/S change to delete designation of the fuel

assembly types allowed in the spent fuel storage racks and new fuel

storage racks is administrative, and does not alter the design and

analysis requirements that ensure storage of fuel in safe

configurations. The existing T/S requirements for maximum

enrichment, reactivity, and spacing of fuel assemblies in the spent

fuel storage racks and new fuel storage racks are not altered by

this change.

Based on the above discussions, design basis accident analyses

affected by these

T/S changes remain valid, and the consequences of an accident

previously evaluated are not significantly increased by these

changes.

Therefore, the probability of occurrence or the consequences of

accidents previously evaluated are not significantly increased.

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

kind of accident from any accident previously evaluated?

The proposed T/S change to allow storage and use of fuel rods

clad with ZIRLO cannot create a new or different kind of accident.

Fuel assemblies with ZIRLO clad fuel rods satisfy the same design

bases as those used for fuel assemblies with Zircaloy-4 clad fuel

rods. The design and performance criteria continue to be met and no

new failure mechanisms have been identified. Since the original

design criteria are met, the ZIRLO clad fuel rods cannot be an

initiator for any new accident. The ZIRLO cladding material offers

improved corrosion resistance and structural integrity. The proposed

changes do not affect the design or operation of any other system or

component in the plant. The safety functions of the other

structures, systems, or components are not changed in any manner,

nor is the reliability of any other structure, system, or component

reduced. The changes do not affect the manner by which the facility

is operated and do not change any other facility design feature,

structure, or system. No new or different types of permanent plant

equipment are installed by this proposed

T/S change. In addition, the use of ZIRLO fuel assemblies does not

involve any alterations to permanent plant equipment or plant

operating procedures that would introduce any new or unique

operational mode or accident precursor.

The proposed T/S change to specify an alternate NRC-approved

methodology used to determine the IFBA requirements for Westinghouse

fuel assemblies stored in the new fuel storage racks ensures that a

conservative methodology is used to verify the licensing basis

reactivity limits are not exceeded. The proposed change does not

affect any permanent plant equipment or plant operating procedures,

and cannot be an initiator of an event.

The proposed T/S change to delete designation of the fuel

assembly types allowed in the spent fuel storage racks and new fuel

storage racks is an administrative

[[Page 67336]]

change only. The proposed change does not affect any permanent plant

equipment or plant operating procedures, and cannot be an initiator

of an event.

Since there is no change to the permanent facility or plant

operating procedures, and the safety functions and reliability of

structures, systems, or components are not affected, the proposed

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

accident from any accident previously evaluated.

Therefore, it is concluded that the change does not create the

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

previously evaluated.

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

of safety?

The proposed T/S change to allow storage and use of fuel rods

clad with ZIRLO does not change the reactor fuel reload design and

safety analysis limits. The use of these fuel assemblies takes into

consideration the core operating conditions allowed in the T/S. For

each cycle reload core, the fuel assembly design and core

configuration are evaluated using NRC-approved reload design

methods, including consideration of the core physics analysis

peaking factors and core average linear heat rate effects. The

design basis and modeling techniques for fuel assemblies with

Zircaloy-4 clad fuel rods remain valid for fuel assemblies with

ZIRLO clad fuel rods. Use of ZIRLO cladding material has no effect

on the criticality analysis for the spent fuel storage racks and the

new fuel storage racks. Furthermore, it has no effect on the

thermal-hydraulic and structural analysis for the spent fuel pool.

Therefore, the design and safety analysis limits specified in the T/

S are maintained with this proposed change.

The proposed T/S change to specify an alternate NRC-approved

methodology used to determine the IFBA requirements for Westinghouse

fuel assemblies stored in the new fuel storage racks ensures that a

conservative methodology is used to verify the licensing basis

reactivity limits are not exceeded. Therefore, the existing T/S

margin for reactivity control in the new fuel storage racks is

maintained by this proposed change.

The proposed T/S change to delete designation of the fuel

assembly types allowed in the spent fuel storage racks and new fuel

storage racks is an administrative change, and does not alter any of

the existing T/S limits governing storage and use of reactor fuel.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,

Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.

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

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: October 16, 1998, as supplemented by

letters dated December 30, 1998, May 10, June 15, July 30, August 2,

11, 16, 19, 27, September 10, and 30, 1999.

Description of amendment request: Associated with a Niagara Mohawk

Power Corporation (NMPC or the licensee) application to convert from

the Curent Technical Specifications (CTS) for the Nine Mile Point

Nuclear Power Station, Unit No. 2, to Improved Technical Specifications

(ITS) as contained in Revision 1 of NUREG-1433, and Revision I of

NUREG-1434, ``Standard Technical Specifications for General Electric

Plants, BWR/4 and BWR/6'' dated April 1995, the licensee proposed to

allow two hydrogen recombiners to be inoperable for up to 7 days

provided that the alternate hydrogen control system is found to be

acceptable to the NRC staff as described below.

CTS 3.6.6.1 ACTION only permits one hydrogen recombiner to be

inoperable. If two hydrogen recombiners are inoperable, CTS 3.0.3 is

entered. CTS 3.6.6.1 ACTION has been modified to incorporate Standard

Technical Specification (STS) 3.6.3.1 ACTION B which allows two

hydrogen recombiners to be inoperable for up to 7 days. The use of STS

3.6.3.1 ACTION B is allowed, as specified in a Bases Reviewer's Note,

provided that the alternate hydrogen control system is found to be

acceptable to the NRC staff. Therefore, the licensee proposed to allow

credit be taken for an alternate hydrogen control system in the event

of both hydrogen recombiners are determined to be inoperable for up to

7 days.

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 the criteria set forth in 10 CFR 50.92, NMPC

has evaluated this proposed Technical Specifications change and

determined it does not represent a significant hazards

consideration. The following is provided in support of this

conclusion.

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change extends the functional test frequency of the

hydrogen recombiner system. The hydrogen recombiners are not

considered as initiators for any previously evaluated accidents.

Therefore, the probability of an accident previously evaluated is

not significantly increased. The proposed change does not impact the

Surveillance Requirement itself nor the way in which the

Surveillance is performed. The proposed change does not affect the

availability of the hydrogen recombiners to mitigate an accident

because of the availability of the redundant hydrogen recombiner.

Furthermore, an historical review of surveillance test results

indicated that all failures identified were unique, non-repetitive,

and not related to any time-based failure modes, and indicated no

evidence of any failures that would invalidate the above

conclusions. Therefore, the proposed change does not involve a

significant increase in the consequences of an accident previously

evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve any design changes, plant

modifications, or changes in plant operation. The system will

continue to function in the same way as before the change. In

addition, the Surveillance Requirement itself and the way the

Surveillance is performed will remain unchanged. Furthermore, a

historical review of surveillance test results indicated no evidence

of any failures that would invalidate the above conclusions.

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

new or different kind of accident from any previously evaluated.

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

of safety?

The design, function, and OPERABILITY requirements for the

hydrogen recombiner system are unchanged with this proposed

revision. Although the proposed change will result in an increase in

the interval between surveillance tests, the impact on hydrogen

recombiner availability is small based on the redundant hydrogen

recombiner, and there is no evidence of any failures that would

impact the availability of the hydrogen recombiners. Therefore, the

assumptions in the licensing basis are not impacted, and the

proposed change does not involve a significant reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

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

NRC Section Chief: Sheri R. Peterson.

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

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: October 25, 1999.

[[Page 67337]]

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to add the Oscillation Power

Range Monitor (OPRM) Upscale function and allow the proposed activation

of the OPRM function of automatically detecting and suppressing reactor

instability conditions. Activation of the OPRM is in response to

Generic Letter 94-02, ``Long-Term Solutions and Upgrade of Interim

Operating Recommendations for Thermal-Hydraulic Instabilities in

Boiling Water Reactors,'' licensee's associated commitment to implement

stability solution Option III as described in Licensing Topical Report

NEDO-31960-A, ``BWR Owners' Group Long-Term Stability Solutions

Licensing Methodology,'' and previous Nine Mile Point Unit 2 (NMP2)

License Amendment 80 dated March 31, 1998. The proposed changes would

add the OPRM as a Reactor Protection Sytem (RPS) Functional Unit,

including operability requirements and surveillance tests.

Specifically, the proposed amendment would revise TS 2.2, ``Limiting

Safety System Settings,'' TS 3/4.3.1, ``Reactor Protection System

Instrumentation,'' TS 3/4.4.1, ``Recirculation System,'' and TS

6.9.1.9, ``Administrative Controls-Core Operating Limits Report.'' The

proposed changes to support activation of the OPRM function are

generally consistent with the changes proposed in Licensing Topical

Report NEDC-32410P-A, ``Nuclear Measurement Analysis and Control Power

Range Neutron Monitor (NUMAC PRNM) Plus Option III Stability Trip

Function,'' Supplement 1, dated November 1997. The licensee's submittal

also provides changes to the associated TS Bases and the TS Index (page

ix).

The proposed changes would be made to NMP2's current TS, as well as

to NMP2's improved TS addressed in a previous notice (64 FR 56518,

October 20, 1999).

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 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 addition of the OPRM Upscale functional unit to TSs involves

a system that is intended to detect the symptoms of instability

events and initiate mitigative actions. The worst case failure of

the system involved would be a failure to initiate mitigative

actions (i.e., scram), but no failure can cause an accident. The

removal of certain RCS [Recirculation System] operational

restrictions is justified with the addition of the OPRM functional

unit which will provide an automatic scram in the event of reactor

instabilities. Therefore, the proposed change will not result in a

significant increase in the probability of any accidents previously

evaluated.

The addition of the OPRM Upscale functional unit to the NMP2 TSs

will permit activation of the OPRM. Activation of the OPRM, together

with the NUMAC-PRNM, provides NMP2 the ability to detect and

suppress reactor instabilities. The existing RPS functional units as

well as other plant equipment will continue to perform their

intended function in the event of an accident. The addition of the

OPRM functional unit fulfills the intended purpose of the TS-

required RCS operational restrictions. Therefore, the proposed

change will not result in a significant increase in the consequences

of any accident previously evaluated.

2. 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 addition of the OPRM Upscale functional unit to the NMP2 TSs

will permit activation of the OPRM. Activation of the OPRM, together

with the NUMAC-PRNM, provides NMP2 the ability to detect and

suppress reactor instabilities. The OPRM is a mitigative system

whose addition as an RPS functional unit will not create the

possibility of a new or different accident or adversely affect

existing RPS functional units. The worst case failure of the systems

involved would be failure to initiate mitigative actions, but no

failure can cause an accident. Except for the activation of the

OPRM, no new plant configurations are created. The OPRM Upscale

functional unit fulfills the intended purpose of the existing TS-

required RCS operational restrictions. Therefore, the proposed

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

accident from any previously evaluated.

3. 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 proposed TS changes will not adversely affect the

performance characteristics of RPS instrumentation nor will it

affect the ability of the subject instrumentation to perform its

intended function.

The addition of the OPRM Upscale functional unit to the NMP2 TSs

will permit activation of the OPRM. Activation of the OPRM, together

with the NUMAC-PRNM, provides NMP2 the ability to detect and

suppress reactor instabilities (stability solution Option III)

thereby meeting the requirements of GDC [General Design Criteria] 10

and 12. The NRC has reviewed and accepted the Option III methodology

described in Licensing Topical Report NEDO-31960-A and concluded

that the solution will provide the intended function. The

surveillance testing and frequencies proposed will assure

reliability of the OPRM Upscale function. The purpose of the

existing TS operational restrictions on the RCS will be met by the

automatic scram feature of the OPRM.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

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

NRC Section Chief: Sheri Peterson.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating

Station, (LGS) Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: October 14, 1999.

Description of amendment request: The proposed amendments, if

approved, would revise the LGS, Units 1 and 2, Technical Specifications

(TSs), Sections 2.2., ``Safety Limits and Limiting Safety System

Settings,'' and 3.0/4.0, ``Limiting Conditions for Operation and

Surveillance Requirements.'' The proposed revisions are required to

support installation of a new Power Range Neutron Monitoring (PRNM)

System and incorporate long-term thermal-hydraulic stability solution

hardware.

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 TS changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

As discussed in the Nuclear Measurement Analysis & Control

(NUMAC) PRNM [Power Range Neutron Monitor] Licensing Topical Report

(LTR), the NUMAC PRNM modification and associated changes to the TS

involve equipment that is designed to detect the symptoms of certain

events or accidents and initiate mitigating actions. The worst case

failure of the equipment involved in the modification is a failure

to initiate mitigating action (scram or rod block), but no failure

can cause an accident. The PRNM replacement system is designed to

perform the same operations as the existing Power Range Monitor

System and meets or exceeds all operational requirements. Therefore,

it is concluded that the probability of an accident

[[Page 67338]]

previously evaluated is not increased as a result of replacing the

existing equipment with the PRNM equipment.

The PRNM System reduces the need for tedious operator actions

during normal conditions and allows the operator to focus more on

overall plant conditions. The automatic self-test and increased

operator information provided with the replacement system are likely

to reduce the burden during off-normal conditions as well. The

replacement equipment qualifications fully envelope the

environmental conditions, including electromagnetic interference, in

the LGS control room.

The replacement equipment has been specifically designed to

assure that it fully meets the response time requirements in the

worst case. As a result, due to statistical variations resulting

from the sampling and update cycles, the response time is typically

faster than required in order to assure that the required response

time is always met. Setpoints are changed only when justified by the

improved equipment performance specifications and by setpoint

calculations which show that safety margins are maintained. There is

no impact to the Control Rod Drop accident analysis because the PRNM

System maintains all existing system functions with a reliability

equal to or better than the existing Power Range Monitor System.

The replacement equipment includes up to 5 LPRM [Local Power

Range Monitor] inputs on a single module compared to one per module

on the current system. Up to 17 LPRM signals are processed through

one preprocessor. The recirculation flow signals are processed in

the same hardware as the LPRM processing. The net effect of these

architectural aspects is that there are some single failures that

can cause a greater loss of ``sub-functionality'' than in the

current system. Other architectural and functional aspects, however,

have an offsetting effect. Redundant power supplies are used so that

a single failure of Reactor Protection System (RPS) AC power has no

effect on the overall PRNM System functions while still resulting in

a half scram as does the current system. Continuous automatic self-

test also assures that if a single failure does occur, it is much

more likely to be detected immediately. The net effect is that from

a total system level, unavailability of the safety-related functions

in the replacement system is equal to or better than the current

Power Range Monitor System.

Based on the extensive and thorough verification and validation

program used in the PRNM design and field operating experience,

common cause failures in software controlled functions are judged to

not be a significant failure mode.

However, in spite of that conclusion, means are provided within

the system to mitigate the effects of such a failure and alert the

operator. Therefore, such a failure, even if it occurred, will not

increase the consequences of a previously evaluated accident.

To reduce the likelihood of common cause failure of software

controlled functions, thorough and careful verification and

validation activities are performed both for the requirements and

the implementing software design. In addition, the software is

designed to limit the loading that external systems or equipment can

place on the system, thus significantly reducing the risk that some

abnormal dynamic condition external to the system can cause system

functional performance problems due to processing ``overload''

(i.e., ``slowing down'' or stopping the processing).

As a conservatism, however, despite these verification and

validation activities, common cause failures of software-controlled

functions due to residual software design faults are assumed to

occur. Both the software and hardware are designed to manage the

consequences of such failure (and also cover potential common cause

hardware failures). Safety outputs are designed to be fail safe by

requiring dynamic update of output modules or data signals, where

failure to update the information is detected by simple receiving

hardware, which, in turn, forces a trip. This aspect covers all but

rather complex failures where the software or hardware executes a

portion of the overall logic but fails to process some portion of

new information (inputs ``freeze'') or some portion of the logic

(outputs ``freeze'').

To help reduce the likelihood of complex failures, a watchdog

timer is used which is updated by a very simple software routine

that in turn monitors the operational cycle time of all tasks in the

system. The software design is such that as long as all tasks are

updated at the design rate, it is likely that software controlled

functions are executing as intended. Conversely, if any task fails

to update at the design rate, that is a strong indication of at

least some unanticipated condition. If such a condition occurs, the

watchdog timer will not be updated, the computer will be

automatically restarted, and the system will detect an abnormal

condition and provide an alarm and trip.

The information available to the operator is at least the same

as with the current system and, in many cases, improved. No actions

are required by the operator to obtain information normally used and

equivalent to that available with the current equipment. However,

the replacement system does provide more directly accessible

information regarding the condition of the equipment, including

automatic self-test, which can aid the operator in diagnosing

unusual situations beyond those defined in the licensing basis.

In summary, the reliability of the new PRNM System and its

ability to detect and mitigate abnormal flux transients have either

remained the same or improved over the existing Power Range Monitor

System. Since these postulated reactivity transients are mitigated

by the new system as effectively and reliability [reliably] as the

existing system, the consequences of these transients have not

changed. Therefore, the proposed TS changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

LGS Modification P00224 uses digital processing with software

(firmware) control for the main signal processing part of the

modification. The remainder of the equipment in the modification

uses conventional equipment similar to the current system (e.g.,

penetrations, cables, interface panels).

The digital equipment has ``control'' processing points and

software-controlled digital processing where as the current system

has analog and discrete component processing. The result is that the

specific failures of hardware and potential software common cause

failures are different from the current system. The effects of

software common cause failure are mitigated by hardware design and

system architecture, but are of a ``different type'' of failure than

those evaluated in the LGS Updated Final Safety Analysis Report

(UFSAR). Therefore, the replacement system may have a malfunction of

a different type from those evaluated in the LGS UFSAR[. . .]

However, when these PRNM failures are evaluated at the system level,

there are no new effects.

LGS Modification P00224 involves equipment that is intended to

detect the symptoms of certain transients and accidents and initiate

mitigating action. The worst case failure of the equipment involved

in the modification is a failure to initiate mitigating action

(scram), but no failure can cause an accident. This is unchanged

from the current system. Software common cause failures could result

in the system failing to perform its safety function, but this

possibility is addressed in Section 1, above. In that case, it might

fail to initiate action to mitigate the consequences of an accident,

but would not cause one. No new system level failure modes are

created with the PRNM System.

Therefore, LGS Modification P00224 does not create the

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

previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in the margin of safety.

The PRNM System response time and operator information is either

maintained or improved over the current Power Range Monitor System.

The PRNM System has improved channel trip accuracy compared to

the current system and meets or exceeds system requirements assumed

in setpoint analysis. The channel response time exceeds the

requirements. The channel indicated accuracy is improved over the

current system and meets or exceeds all of the system requirements.

The PRNM System was developed to detect the presence of thermal-

hydraulic instabilities and automatically initiate the necessary

corrective actions to suppress the oscillations prior to violating

the Minimum Critical Power Ratio (MCPR) Safety Limit. The NRC has

reviewed and approved the PRNM Licensing Topical Report (LTR)

concluding that the PRNM System will provide the intended

protection.

Therefore, LGS Modification P00224 does not result in a

significant reduction in the margin of safety.

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

this review, it appears that the three

[[Page 67339]]

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.

Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101.

NRC Section Chief: James W. Clifford.

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

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

Date of amendment request: September 9, 1996, as supplemented on

June 6, 1997, and June 7, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 Technical Specifications (TSs) proposes to revise TS

Section 6 to delete requirements for Plant Operating Review Committee

review of the fire protection program and implementing procedures.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the Indian Point 3 plant in accordance with the

proposed amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92, since it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes delete the Plant Operating Review Committee

(PORC) review of changes to the fire protection program and

implementing procedures. The changes do not introduce any new modes

of plant operation, make any physical changes, or alter any

operational setpoints. Therefore, the changes do not degrade the

performance of any safety system assumed to function in the accident

analysis. Consequently, there is no effect on the probability or

consequences of an accident.

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

from those previously evaluated.

No physical changes to the plant or changes to equipment

operating procedures are proposed. The changes are administrative

and will not have any direct effect on equipment important to

safety. Therefore the changes cannot create the possibility of a new

or different kind of accident.

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

Adequacy of the fire protection program and implementing

procedures is assured by the fire protection license condition, the

procedure review and approval process implemented by Amendment 159,

the provisions of 10 CFR 50.59, and inspections and audits performed

under the cognizance of the SRC [Safety Review Committee].

Consequently, deleting PORC's responsibility for review of the fire

protection program and implementing procedure will not degrade the

fire protection program. Therefore, the proposed changes do not

involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Section Chief: Sheri R. Peterson.

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

362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego

County, California

Date of amendment request: November 8, 1999 (PCN 454).

Description of amendment requests: The licensee proposed to revise

Surveillance Requirement (SR) 3.8.1.18 of Technical Specification (TS)

3.8.1, ``A.C. Sources-Operating.'' Currently, SR 3.8.1.18 reads: Verify

interval between each sequenced load block is within plus or minus 10%

of design interval for each emergency and shutdown load programmed time

interval load sequence. The licensee proposed to revise the SR to read:

Verify the timing of each sequenced load block is within its timer

setting plus or minus 10% or plus or minus 2.5 seconds, whichever is

greater, with the exception of the 5 second load group which is minus

0.5, plus 2.5 seconds, for each programmed time interval load sequence.

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. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of any accident previously evaluated?

Response: No.

The proposed change would expand the current surveillance

acceptance criteria to more accurately reflect the characteristics

of the installed plant equipment. The diesel generators (DG's) have

sufficient capacity to maintain adequate voltage and frequency

during load sequencing with the expanded tolerance. The overall

Engineered Safety Features (ESF) response times in the Technical

Specifications and safety analyses are maintained even though the

timer tolerance is increased. Therefore, the consequences of any

accident previously evaluated are not increased. The DG load

sequence timers are not of themselves a credible initiator of any

accident, so the probability of an accident has not been increased.

The timers will function acceptably to support the equipment needed

for accident mitigation, so the consequences of an accident are not

increased. Therefore, the probability or consequences of any

accident previously evaluated are not increased.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

This amendment request does not involve any change to plant

equipment or operation. In the event of a loss of preferred power,

the ESF electrical loads are automatically connected to the DG's in

sufficient time to provide for safe reactor shutdown and to mitigate

the consequences of a Design Basis Accident such as a loss of

coolant accident. Increasing the timer tolerance will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

This amendment does not change the manner in which safety

limits, limiting safety settings, or limiting conditions for

operations are determined. The actual response times have not been

altered by this amendment. Therefore, operation of equipment will

not be affected. Accordingly, this 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 10 CFR 50.92(c) are

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

amendment requests involve no significant hazards consideration.

Attorney for licensee: Douglas K. Porter, Esquire, Southern

California Edison Company, 2244 Walnut Grove Avenue, Rosemead,

California 91770.

NRC Section Chief: Stephen Dembek.

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

362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego

County, California.

Date of amendment request: November 12, 1999 (PCN 505).

Description of amendment requests: The licensee proposed to revise

Technical Specification (TS) 5.5.2.13, ``Diesel Fuel Oil Testing

Program.'' Specifically, the following changes are proposed:

1. The at least once per 92 days test is deleted for water and

sediment,

[[Page 67340]]

American Petroleum Institute (API) gravity or an absolute specific

gravity, and kinematic viscosity for the diesel fuel oil in the

Emergency Diesel Generator fuel oil storage tanks. The requirement to

test these properties prior to addition of new fuel to the storage tank

remains unchanged.

2. A requirement is added to test new fuel oil prior to addition to

the storage tank to verify that the flash point is within limits.

3. A requirement is added to test new fuel oil within 31 days of

delivery for ``other properties for ASTM [American Society for Testing

and Materials] 2D fuel.''

4. The acceptance criteria for the properties listed, with the

exception of the particulate criterion, are replaced with the phrase

``within limits.'' The statement which requires sampling in accordance

with ASTM-D4057-81 is deleted. Acceptance criteria and reference to the

applicable standard for sampling are currently provided in the Bases

for Surveillance Requirement 3.8.3.3.

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) Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

This change is an administrative change to make Technical

Specification (TS) 5.5.2.13, ``Diesel Fuel Oil Testing Program,''

consistent with the existing Bases for Surveillance Requirement (SR)

3.8.3.3. The specific changes are:

1. The at least once per 92 days diesel fuel oil test is deleted

for water and sediment, American Petroleum Institute (API) gravity

or an absolute specific gravity, and kinematic viscosity. The

requirement to test these properties prior to addition of new fuel

to the storage tank remains unchanged.

2. A requirement is added to test new fuel oil prior to addition

to the storage tank to verify that the flash point is within limits.

3. A requirement is added to test new fuel oil within 31 days of

delivery for ``other properties for ASTM 2D fuel.''

4. The acceptance criteria for the properties listed, with the

exception of the particulate content, are replaced with the phrase

``within limits.'' Acceptance criteria are currently provided in the

Bases for Surveillance Requirement 3.8.3.3.

These changes are all consistent with the existing Bases for SR

3.8.3.3 and NUREG 1432.

Therefore, this change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

This change is an administrative change to make TS 5.5.2.13,

``Diesel Fuel Oil Testing Program,'' consistent with the existing

Bases for Surveillance Requirement 3.8.3.3.

Therefore, this proposed change will not create the possibility

of a new or different kind of accident from any accident that has

been previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

This change is an administrative change to make TS 5.5.2.13,

``Diesel Fuel Oil Testing Program,'' consistent with the existing

Bases for Surveillance Requirement 3.8.3.3.

Therefore, there will be no significant reduction in a margin of

safety as a result of this change.

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

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

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

amendment requests involve no significant hazards consideration.

Attorney for licensee: Douglas K. Porter, Esquire, Southern

California Edison Company, 2244 Walnut Grove Avenue, Rosemead,

California 91770.

NRC Section Chief: Stephen Dembek.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424

and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2,

Burke County, Georgia

Date of amendment request: April 19, 1999, as supplemented by

letter dated November 1, 1999.

Description of amendment request: The proposed change would revise

Surveillance Requirement (SR) 3.3.5.2 and associated Bases to allow the

loss of voltage and degraded voltage trip setpoints to be treated as

nominal values in the same manner as the trip setpoints for the Reactor

Trip System (RTS) and Engineered Safety Feature Actuation System

(ESFAS) instrumentation. The November 1, 1999, letter removes a note

proposed in the April 19, 1999, amendment request. This revision does

not change the scope of the April 19, 1999, application and the initial

proposed no significant hazards consideration.

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 proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

No. The proposed change affects only the presentation of the

trip setpoints for loss of voltage and degraded voltage in SR

3.3.5.2 in the VEGP Units 1 and 2 TS [Technical Specifications]. The

calibration of the channels whose setpoints are specified in SR

3.3.5.2 will continue to be performed in a manner consistent with

the setpoint methodology used to determine the trip setpoints. There

will be no adverse effect on the ability of those channels to

perform their safety functions as assumed in the safety analyses.

Since there will be no adverse effect on the trip setpoints or the

instrumentation associated with those trip setpoints, there will be

no increase in the probability of any accident previously evaluated.

Similarly, since the ability of the instrumentation to perform its

safety function is not adversely affected, there will be no increase

in the consequences of any accident previously evaluated.

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

different kind of accident from any accident previously evaluated?

No. The proposed change affects only the presentation of the

trip setpoint requirements of SR 3.3.5.2. Plant operation will not

be changed, and the response of safety related equipment as assumed

in the accident analyses would not be adversely affected. Therefore,

the proposed change does not involve a new or different kind of

accident than any previously evaluated.

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

margin of safety?

No. As described above, the loss of voltage and degraded voltage

instrumentation will remain capable of performing its safety

function as assumed in the accident analyses. The treatment of trip

setpoints as nominal values is consistent with the methodology used

to establish those setpoints. As such, margin is not affected by the

proposed change.

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.

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

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

Georgia 30308-2216.

NRC Section Chief: Richard L. Emch, Jr.

[[Page 67341]]

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: September 8, 1999, as supplemented by

letter dated November 9, 1999. The September 8, 1999, application was

originally noticed in the Federal Register on November 3, 1999 (64 FR

59806).

Description of amendment request: The proposed amendments would

revise Technical Specification 3/4.8.1, ``A.C. Sources, Operating,''

and associated Bases, by relocating the 18-month surveillance to

subject the standby diesel generator to inspections, in accordance with

procedures prepared in conjunction with its manufacturer's

recommendations, to the Technical Requirements Manual.

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 moves the requirement to perform

manufacturer's recommended inspections of the Standby Diesel

Generators from the Technical Specifications to the Technical

Requirements Manual (TRM). The change does not result in any

hardware or operating procedure changes. The requirement being

removed from the Technical Specifications is not the initiator of

any analyzed event. The TRM is maintained using the provisions of 10

CFR 50.59. Since any changes will be evaluated per 10 CFR 50.59, no

significant increase in the probability or consequences of an

accident previously evaluated will be allowed without prior NRC

approval. Therefore, the changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change moves the requirement to perform

manufacturer's recommended inspections of the Standby Diesel

Generators from the Technical Specifications to the TRM. The change

does not alter the plant configuration (no new or different type of

equipment will be installed) or make changes in methods governing

normal plant operation. The change does not impose different

requirements. The change does not alter assumptions made in the

safety analysis and licensing basis. Therefore, the change will not

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

any accident previously evaluated.

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

of safety?

The proposed change moves the requirement to perform

manufacturer's recommended inspections of the Standby Diesel

Generators from the Technical Specifications to the TRM. The change

does not reduce the margin of safety since the location of details

has no impact on any safety analysis assumptions. In addition, the

requirement being transposed from the Technical Specification to the

TRM is the same as the existing Technical Specification. Also, the

TRM is maintained using the provisions of 10 CFR 50.59. Since any

changes will be evaluated per 10 CFR 50.59, no significant reduction

in a margin of safety will be allowed without prior NRC approval.

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

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

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

request for amendments involves no significant hazards consideration.

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

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Section Chief: Robert A. Gramm.

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.

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

C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment requests: November 5, 1999.

Description of amendment requests: The proposed license amendments

would revise Technical Specification (T/S) Surveillance Requirement

4.5.1.c to require verification that power is removed from each

emergency core cooling system accumulator isolation valve operator

instead of verification that each accumulator isolation valve breaker

is removed from the circuit. In addition, the proposed license

amendments would revise T/S 3.5.1 to change ``pressurizer pressure'' to

``reactor coolant system pressure'' in the applicability and action

statement requirements. The Bases for T/S 3/4.5.1 will also be revised

to reflect both changes. Additionally, administrative changes are

proposed to the page format.

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 of occurrence or consequences of an accident previously

evaluated?

The ECCS [emergency core cooling system] accumulators are used

to mitigate the consequences of an accident after the event has

occurred and do not initiate any accident previously evaluated.

Demonstrating how power is removed from the valve operator does not

initiate an accident. Inadvertently closing the valves cannot

initiate an accident. Therefore, there is no significant increase in

the probability of occurrence of an accident previously evaluated.

The ECCS accumulators will still perform their function of

injecting borated water into the reactor coolant loops following a

large break loss-of-coolant accident, as described in Section 14.3.1

of the Updated Final Safety Analysis Report (UFSAR). A spurious

closure of an accumulator outlet isolation valve is not a credible

event. Performing T/S Surveillance Requirement 4.5.1.c provides

assurance that one of the two actions required for spurious closure

of the valve is precluded. The proposed change to the surveillance

continues to provide assurance that power will be removed from each

accumulator isolation valve operator so that the valves remain open.

The consequences of accidents previously evaluated remained bounded

because the accumulators will still function as assumed in the UFSAR

accident analysis. Therefore, there is no significant increase in

the consequences of any accident previously evaluated.

Changing ``pressurizer pressure'' to ``RCS [reactor coolant

system] pressure'' has no significant effect on the applicability of

the T/S requirements. RCS pressure and pressurizer pressure

instrumentation measure a similar parameter in the primary coolant

system. Since the RCS is a closed-loop fluid system, pressure

instruments should indicate approximately the same value. There is

no significant difference between the instrument readings because

they are corrected for range, height, and accuracy. There is no

significant change in the margin of pressure between when the

accumulators are required to be aligned at 1000 psig and the upper

limit specified in T/S 3.5.1.d of 658 psig.

The proposed format changes are administrative and have no

impact on plant operation.

Therefore, the proposed changes do not increase the probability

of occurrence or

[[Page 67342]]

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes to T/S 3/4.5.1 and the associated Bases do

not involve any physical changes to the plant, but do change the way

the plant is operated by changing the method for ensuring spurious

closure of the accumulator isolation valve will not occur. The

proposed change to T/S Surveillance Requirement 4.5.1.c does not

create any new operator actions. The position of the accumulator

isolation valve remains open in Modes 1, 2, and 3 with RCS pressure

greater than 1000 psig, which meets its design safety function. The

proposed change does not increase the possibility of the accumulator

valve repositioning. In order for repositioning to happen, the

operator must close the molded-case circuit breaker coupled with

either an active single failure or deliberate operator action in the

control room. The proposed change of verifying that power is removed

from the accumulator isolation valve provides the same level of

protection. Two positive actions are required for the accumulator

isolation valve to reposition.

The proposed format changes are administrative and have no

impact on plant operation.

Therefore, the proposed changes do 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?

T/S Surveillance Requirement 4.5.1.c provides requirements that

ensure that a single action will not cause an inadvertent closure of

the accumulator isolation valves. The proposed change continues to

ensure that two positive actions, an operator action to restore the

breaker and a single failure, are required for valve closure.

Changing ``pressurizer pressure'' to ``RCS pressure'' does not

impact operation of the accumulators. The proposed changes do not

impact the nitrogen cover pressure as stated in T/S 3.5.1.c. The

accumulators would not be expected to inject borated water until RCS

pressure lowers to 658 psig (the upper limit specified in T/S

3.5.1.d). The change does not affect when this would occur after an

accident. Therefore, changing ``pressurizer pressure'' to ``RCS

pressure'' has no impact on plant operation.

The proposed format changes are administrative and have no

impact on plant operation.

Therefore, there is no significant reduction in the margin of

safety.

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

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

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

amendment requests involves no significant hazards consideration.

Attorney for licensee: David W Jenkins, Esq., 500 Circle Drive,

Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.

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 electronically from the ADAMS Public

Library component on the NRC Web site, http://www.nrc.gov (the

Electronic Reading Room).

Arizona Public Service Company, et al., Docket No. STN 50-528, Palo

Verde Nuclear Generating Station, Unit No. 1, Maricopa County, Arizona

Date of application for amendment: October 8, 1999, as supplemented

October 29, 1999.

Brief description of amendment: The amendment revises Surveillance

Requirement 3.8.4.8 of Technical Specification 3.8.4, to allow the

licensee to forego the performance of this surveillance until entry

into MODE 4 coming out of the ninth refueling outage for Unit 1.

Date of issuance: November 19, 1999.

Effective date: November 19, 1999.

Amendment No.: 121.

Facility Operating License No. NPF-41: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 19, 1999 (64 FR

56369).

The October 29, 1999, supplement provided clarifying information

that was within the scope of the original Federal Register notice and

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

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 19, 1999.

No significant hazards consideration comments received: No

Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station,

Unit 1, LaSalle County, Illinois

Date of application for amendment: July 7, 1999, as supplemented on

October 14, 1999.

Brief description of amendment: The amendment revised Section 2.1

of the Technical Specifications to reflect a change in the Minimum

Critical Power Ratio.

Date of issuance: November 9, 1999.

Effective date: Immediately, to be implemented prior to the startup

of Cycle 9.

Amendment No.: 137.

Facility Operating License No. NPF-11: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 11, 1999.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 9, 1999.

No significant hazards consideration comments received: No.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power

Station, Unit 2, Shippingport, Pennsylvania

Date of application for amendment: January 29, 1998, as

supplemented by letters dated November 9, 1998, and June 14, 1999.

Brief description of amendment: This amendment authorized changes

to the Beaver Valley Power Station, Unit No. 2 (BVPS-2) Updated Final

Safety Analysis Report (UFSAR). The amendment authorizes changes to the

UFSAR to reflect revisions to the radiological dose calculations for

the locked rotor accident analysis. This revision of the calculation

was performed in order to incorporate more conservative

[[Page 67343]]

assumptions than those used in the previous analysis for a postulated

locked rotor event.

These changes are not the result of hardware changes to the plant

or any change in operating practices. They reflect revised analysis

results only and allow revision of the licensing basis to reflect

conservative assumptions used in the revised analyses.

The June 14, 1999, letter withdrew a portion of the amendment which

would have revised the UFSAR description of the small-break loss-of-

coolant accident radiological consequences.

Date of issuance: November 18, 1999.

Effective date: As of the date of issuance.

Amendment No: 103.

Facility Operating License No. NPF-73. Amendment approved changes

to the UFSAR.

Date of initial notice in Federal Register: March 11, 1998 (63 FR

11919).

The November 9, 1998, and June 14, 1999, letters provided

clarifying information that did not change the initial proposed no

significant hazards consideration determination or expand the amendment

beyond the scope of the initial notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 18, 1999.

No significant hazards consideration comments received: No.

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

No. 2, Pope County, Arkansas

Date of application for amendment: July 29, 1999, as supplemented

by letters dated August 6, 1999, October 14, 1999, and October 26,

1999.

Brief description of amendment: The proposed change to the Arkansas

Nuclear One, Unit No. 2 Technical Specifications would allow the

performance of a special inspection of the steam generator tubes during

an upcoming mid-cycle outage. This mid-cycle outage is planned for the

purpose of performing inspections in selected areas of the steam

generator tube bundle where previous inspections have revealed tube

degradation. The proposed change would limit the initial inspection

scope to these identified areas and includes scope expansion criteria

to address unexpected results.

Date of issuance: November 5, 1999.

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

within 30 days from the date of issuance.

Amendment No.: 210.

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

Technical Specifications.

Date of initial notice in Federal Register: October 6, 1999 (64 FR

54375).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 5, 1999.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment: May 6, 1999.

Brief description of amendment: The amendment incorporates the

Technical Specification changes necessary for redefining the minimum

critical power ratio safety limit for Cycle 11 operation with a mixed

core of Siemens Power Corporation fuel and General Electric fuel.

Date of issuance: November 17, 1999.

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

within 30 days of issuance.

Amendment No: 140.

Facility Operating License No. NPF-29: The amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: August 25, 1999 (64 FR

46434).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 17, 1999.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit 1, Ottawa County, Ohio.

Date of application for amendment: July 26, 1999.

Brief description of amendment: This amendment--

(1) Relocates the requirements in TS 3/4.3.3.2, ``Instrumentation--

Incore Detectors,'' TS 3/4.3.3.9, ``Instrumentation--Waste Gas System

Oxygen Monitor,'' and TS 3/4.4.4.7, ``Reactor Coolant System--

Chemistry,'' to the Davis-Besse Nuclear Power Station (DBNPS) Updated

Safety Analysis Report (USAR) Technical Requirements Manual (TRM);

(2) Revises TS 3/4.11.2, ``Radioactive Effluents--Explosive Gas

Mixture,'' to reflect the relocation of TS 3/4.3.3.9;

(3) Revises the requirements of TS 3/4.4.6.1, ``Reactor Coolant

System Leakage--Leakage Detection Systems,'' to require one monitor

(gaseous or particulate) of the containment atmosphere radioactivity

monitoring systems to be operable, rather than requiring both systems

to be operable simultaneously; and

(4) Revises TS 3/4.3.3.1, ``Radiation Monitoring Instrumentation,''

to be consistent with the revision to TS 3/4.4.6.1.

Date of issuance: November 16, 1999

Effective date: November 16, 1999.

Amendment No.: 234.

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

Technical Specifications.

Date of initial notice in Federal Register: August 25, 1999 (64 FR

46436).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 16, 1999

No significant hazards consideration comments received: No

NASA Aeronautics Space Administration (NASA), Docket No. 50-30, NASA

Test Reactor, Erie County, Ohio

Date of application for amendment: March 25, 1999, as supplemented

on August 10, 1999.

Brief description of amendment: This amendment changes Lewis

Research Center (LeRC) to Glenn Research Center (GRC).

Date of issuance: November 16, 1999.

Effective Date: November 16, 1999.

Amendment No: 10.

Facility License No. TR-3: The amendment changes facility name.

Date of initial notice in Federal Register: October 6, 1999 (64 FR

54377).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 16, 1999.

No significant hazards consideration comments received: No.

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

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

Date of application for amendment: November 16, 1998, as

supplemented June 21, 1999.

Brief description of amendment: Amendment changes Technical

Specifications to limit reactor power oscillations during a reactor

trip and allows operation in the Extended Load Line Limit Analysis

region of the power/flow operating curve.

Date of issuance: September 21, 1999.

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

30 days.

Amendment No.: 168.

[[Page 67344]]

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

Technical Specifications.

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

FR 71968) as corrected January 27, 1999 (64 FR 4148).

The June 21, 1999, letter provided supporting information that did

not change the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 21, 1999.

No significant hazards consideration comments received: No.

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

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

Date of amendment request: September 29, 1998, as supplemented by

letters dated March 8 and April 7, 1999.

Description of amendment request: To revise Facility Operating

License No. NPF-86 to reflect the transfer of the license, to the

extent held by Montaup Electric Company, to Little Bay Power

Corporation.

Date of issuance: November 19, 1999.

Effective date: As of its date of issuance, and shall be

implemented within 30 days.

Amendment No.: 65.

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

License.

Date of initial notice in Federal Register: December 14, 1998 (63

FR 68801). The March 8 and April 7, 1999 supplements provided

clarifying information and did not change the staff's proposed no

significant hazards determination. The Commission received comments

which were addressed in the staff's Safety Evaluation dated August 3,

1999. The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 3, 1999.

No significant hazards consideration comments received: Yes.

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

Nuclear Power Station, Unit No. 1, New London County, Connecticut

Date of application for amendments: April 19, 1999, as supplemented

August 25, October 14, and November 3, 1999.

Brief description of amendments: The amendment deletes most of the

current Technical Specifications to implement the Permanently Defueled

Technical Specification. Portions of the April 19, 1999, request

related to fuel storage pool water level, crane operability, and crane

travel with a spent fuel cask will be addressed at a later date.

Date of issuance: November 9, 1999.

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

within 90 days from the date of issuance.

Amendment No.: 106.

Facility Operating License No. DPR-21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 30, 1999 (64 FR

35208).

The August 25, 1999, letter provided clarifying information that

did not change the scope of the April 19, 1999, application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 9, 1999.

No significant hazards consideration comments received: No

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

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

Date of application for amendment: May 7, 1998, as supplemented

January 22, 1999.

Brief description of amendment: The amendment revises the licensing

basis to address the addition of the dose from the Refueling Water

Storage Tank back leakage into the design basis loss-of-coolant

accident analysis and Chapter 15 of the Final Safety Analysis Report.

Date of issuance: November 4, 1999.

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

implemented within 60 days.

Amendment No.: 176.

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

Technical Specifications.

Date of initial notice in Federal Register: July 1, 1998 (63 FR

35991). The January 22, 1999, supplement provided clarifying

information that did not change the staff's initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 4, 1999.

No significant hazards consideration comments received: No

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

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

Date of application for amendment: August 5, 1999.

Brief description of amendment: The amendment corrects editorial

errors in the Technical Specifications Sections 3.8.3.2, 4.6.2.1,

4.8.1.1, and 4.9.12. The amendment also corrects minor editorial and

reference errors in Bases Sections B 3/4.3.2, B 3/4.4.11, B 3/4.6.1.2,

and B 3/4.8.4.

Date of issuance: November 15, 1999.

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

within 30 days from the date of issuance.

Amendment No.: 177.

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

Technical Specifications.

Date of initial notice in Federal Register: September 8, 1999 (64

FR 48858).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 15, 1999.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of application for amendments: December 29, 1998, as

supplemented by letters dated July 30 and October 12, 1999.

Brief description of amendments: The amendments revise Technical

Specifications (TS) 6.9.1.8, ``Core Operating Limits Report,'' of the

current TSs and TS 5.6 of the improved TSs, to allow the use of NRC

approved addenda to WCAP-10054-P-A, ``Westinghouse Small Break ECCS

Evaluation Model Using NOTRUMP Code,'' August 1985, to determine core

operating limits. The improved TSs were issued in Amendment Nos. 135

for Diablo Canyon Power Plant, Units 1 and 2 dated May 28, 1999, but

have not yet been implemented.

Date of issuance: November 15, 1999.

Effective date: November 15, 1999, and shall be implemented within

90 days from the date of issuance.

Amendment Nos.: Unit 1--136; Unit 2-136.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 21, 1999 (64 FR

19562). The July 30 and October 12, 1999, supplemental letters provided

additional clarifying information and did not change the staff's

initial no significant hazards consideration determination. The

Commission's related evaluation of the amendments is

[[Page 67345]]

contained in a Safety Evaluation dated November 15, 1999.

No significant hazards consideration comments received: No.

PECO Energy Company, Docket No. 50-352, Limerick Generating Station,

Unit 1, Montgomery County, Pennsylvania.

Date of amendment request: January 12, 1999, as supplemented

January 29, March 10, and September 20, 1999.

Description of amendment request: This amendment revised Technical

Specifications (TSs) Section 3/4.4.2, ``Safety/Relief Valves,'' and TS

Bases Sections B 3/4.4.2, B 3/4.5.1 and B 3/4.5.2 to increase the

allowable as-found main steam safety relief valve (SRV) code safety

function lift setpoint tolerance from plus or minus 1% to plus or minus

3%. Also, the required number of operable SRVs in operational

conditions 1, 2, and 3 will be increased from 11 to 12.

Date of issuance: November 10, 1999.

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

prior to completion of the spring 2000 refueling outage for Limerick

Generating Station, Unit 1.

Amendment No: 137.

Facility Operating License No. NPF-39. The amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: February 24, 1999 (64

FR 9194).

The January 29, March 10, and September 20, 1999, letters provided

clarifying information that did not change the initial proposed no

significant hazards consideration determination or expand the scope of

the original Federal Register notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 10, 1999.

No significant hazards consideration comments received: No.

PECO Energy Company, Docket No. 50-352, Limerick Generating Station,

Unit 1, Montgomery County, Pennsylvania.

Date of application for amendment: June 7, 1999.

Brief description of amendment: The amendment revised the technical

specifications (TSs) to reflect the permanent deactivation in the

closed position of the ``wet'' instrument reference leg isolation valve

HV-61-102. Specifically, TS Table 3.6.3.1, ``Primary Containment

Isolation Valve,'' and its associated notations were revised to reflect

this current plant configuration.

Date of issuance: November 18, 1999.

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

within 30 days.

Amendment No.: 138.

Facility Operating License No. NPF-39. This amendment revised the

TSs.

Date of initial notice in Federal Register: October 6, 1999 (64 FR

54380).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 18, 1999.

No significant hazards consideration comments received: No.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic City Electric Company, Docket

Nos. 50-277 and 278, Peach Bottom Atomic Power Station, Unit Nos. 2 and

3, York County, Pennsylvania

Date of application for amendments: December 24, 1998, as

supplemented May 25 and September 27, 1999.

Brief description of amendments: These amendments revise Technical

Specification (TS) Table 3.3.8.1-1 related to loss of power

instrumentation set points and limits of allowable values for the 4 kV

emergency buses.

Date of issuance: November 16, 1999.

Effective date: These license amendments are effective as of their

date of issuance. Phase 1 applies to Functions 2 and 3 in TS Table

3.3.8.1-1 and shall be implemented within 30 days of the date of

issuance of the amendment. Phase 2 applies to Functions 4 and 5 in TS

Table 3.3.8.1-1 and shall be implemented no later than March 1, 2000.

Note (a) shall be implemented within 30 days of the date of issuance of

the amendment and shall be voided upon completion of modification 96-

01511, but no later than March 1, 2000.

Amendments Nos.: 230 and 235.

Facility Operating License Nos. DPR-44 and DPR-56: The amendments

revised the Technical Specifications. The May 25 and September 27,

1999, letters provided clarifying information that did not change the

initial proposed no significant hazards consideration.

Date of initial notice in Federal Register: May 5, 1999 (64 FR

24199).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 16, 1999.

No significant hazards consideration comments received: No.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

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

supplemented July 23, 1998, December 3, 1998, February 25, 1999, and

September 29, 1999.

Brief description of amendment: The amendment revises Technical

Specifications to permit use of additional spent fuel storage racks.

Date of issuance: November 10, 1999.

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

30 days.

Amendment No.: 256.

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

Technical Specifications.

Date of initial notice in Federal Register: August 24, 1998 (63 FR

45096).

The July 23, 1998, December 3, 1998, February 25, 1999, and

September 29, 1999, applications provided supplemental information that

did not affect the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 10, 1999.

No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of application for amendments: November 14, 1997, as

supplemented on August 25, 1999.

Brief description of amendments: The amendments revise the TSs to

make administrative and editorial changes to correct errors in the TSs

that have either existed since initial issuance or were introduced

during subsequent changes. In addition, surveillance requirements are

added that should have been incorporated within the TSs when the

applicable amendment to the TSs was approved by the NRC.

Date of issuance: November 2, 1999.

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

within 60 days.

Amendment Nos.: 225 and 206.

Facility Operating License Nos. DPR-70 and DPR-75: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 17, 1997 (63

FR 66141). The August 25, 1999, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination.

[[Page 67346]]

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 2, 1999.

No significant hazards consideration comments received: No.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco

Nuclear Generating Station, Sacramento County, California

Date of application for amendments: March 18, 1996, as supplemented

April 28, 1997, and February 16, 1999.

Brief description of amendment: The amendment authorizes changes to

the design-basis accident analysis (postulated cask drop accident) to

be incorporated into the Defueled Safety Analysis Report (DSAR) and

revises the Permanently Defueled Technical Specifications to reflect

the changes to the cask drop analysis.

Date of issuance: November 12, 1999.

Effective date: November 12, 1999, with the Technical

Specifications to be implemented within 30 days. Implementation also

includes incorporation of the changes into the DSAR at the next update

of the DSAR in accordance with the schedule in 10 CFR 50.71(e).

Amendment No.: 127.

Facility Operating License No. DPR-54: The amendment revised the

Technical Specifications and the Defueled Safety Analysis Report.

Date of initial notice in Federal Register: August 25, 1999 (64 FR

46442).

The April 28, 1997, and February 16, 1999, supplements provided

additional clarifying information that was within the scope of the

original Federal Register notice and did not change the staff's initial

proposed no significant hazards consideration determination. The

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

Safety Evaluation dated November 12, 1999.

No significant hazards consideration comments received: No.

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

362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego

County, California

Date of application for amendments: October 20, 1998 (PCN 485), as

supplemented August 13, 1999.

Brief description of amendments: The amendments revise Technical

Specification 3.3.9 by adding a surveillance requirement for response

time testing for the control room isolation signal.

Date of issuance: November 15, 1999.

Effective date: November 15, 1999, to be implemented within 30 days

of issuance.

Amendment Nos.: Unit 2--160; Unit 3--151.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 12, 1999 (64 FR

55311).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 15, 1999.

No significant hazards consideration comments received: No.

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: May 7, 1998, as supplemented by letters

dated May 20, June 16, September 30, October 20, and October 21, 1999.

Brief description of amendments: The amendments changed the

Technical Specifications (TSs) to reflect reactor coolant system flow

differences between the existing Model E and replacement Model

94 steam generators (SGs) by adding a new flow rate

requirement to TS 3.2.5, Departure from Nucleate Boiling (DNB)

Parameters, that is applicable to the Model 94 SGs. Related

changes to Bases 3/4.2.5, DNB Parameters, were also made. The licensee

withdrew all changes proposed in the May 7, 1998, application that were

superseded by the previously approved amendments 115/103 dated

September 2, 1999.

Date of issuance: November 8, 1999.

Effective date: November 8, 1999.

Amendment Nos.: Unit 1--117; Unit 2--105.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 1, 1998 (63 FR

35996).

The May 20, June 16, September 30, October 20, and October 21,

1999, supplements provided additional clarifying information. The

September 30, 1999, supplement also provided updated TS pages. This

information was within the scope of the original application and

Federal Register notice and did not change the staff's initial proposed

no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 8, 1999.

No significant hazards consideration comments received: No.

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: August 31, 1998, as supplemented by

letters dated April 19, August 18, and October 21, 1999.

Brief description of amendments: The amendments revised Technical

Specification 3/4.4.9.3 by revising the cold overpressure mitigation

curve to accommodate the replacement steam generators and by adding two

surveillances (for the centrifugal charging pumps and the emergency

core cooling system accumulators) to ensure the operability of the cold

overpressure mitigation system.

Date of issuance: November 9, 1999.

Effective date: November 9, 1999, to be implemented within 30 days.

Amendment Nos.: Unit 1--118; Unit 2--106.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: September 8, 1999 (64

FR 48867).

The October 21, 1999, supplement provided a revised implementation

date. This information was within the scope of the original application

and Federal Register notice and did not change the staff's initial no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 9, 1999.

No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and

Final Determination of No Significant Hazards Consideration and

Opportunity for a Hearing (Exigent Public Announcement or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

Energy Act of 1954, as amended (the Act), and the Commission's rules

and regulations. The Commission has made appropriate findings as

required by the Act and the Commission's rules and regulations in 10

CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the

date

[[Page 67347]]

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

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

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

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

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

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

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

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

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

transcribed as appropriate and the licensee has been informed of the

public comments.

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

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

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

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

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

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

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

Commission may provide an opportunity for public comment. If comments

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

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

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

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

significant hazards consideration is involved.

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

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

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

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

application for amendment, (2) the amendment to Facility Operating

License, and (3) the Commission's related letter, Safety Evaluation

and/or Environmental Assessment, as indicated. All of these items are

available for public inspection at the Commission's Public Document

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

electronically from the ADAMS Public Library component on the NRC Web

site, http://www.nrc.gov (the Electronic Reading Room).

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By January 3, 2000, 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 electronically from the ADAMS Public Library

component on the NRC Web site, http://www.nrc.gov (the Electronic

Reading Room). If a request for a hearing or petition for leave to

intervene is filed by the above date, the Commission or an Atomic

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

Chairman of the Atomic Safety and Licensing Board Panel, will rule on

the request and/or petition; and the Secretary or the designated Atomic

Safety and Licensing Board will issue a notice of a hearing or an

appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has

[[Page 67348]]

made a final determination that the amendment involves no significant

hazards consideration, if a hearing is requested, it will not stay the

effectiveness of the amendment. Any hearing held would take place while

the amendment is in effect.

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

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

Commission, Washington, DC 20555-0001, 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 the

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

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

County, Michigan

Date of amendment request: October 29, 1999, as supplemented

November 2, 1999.

Description of amendment request: The amendment revises the

Technical Specification administrative controls regarding the

containment leak rate testing program and the core operating limits

report. These changes are necessary to reflect changes in the accident

analyses and core design methodologies for the next operating cycle.

Date of issuance: November 15, 1999.

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

within 30 days.

Amendment No.: 188.

Facility Operating License No. DPR-20: Amendment revises the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes. The NRC published a public

notice of the proposed amendment, issued a proposed finding of no

significant hazards consideration, and requested that any comments on

the proposed no significant hazards consideration be provided to the

staff by close of business November 12, 1999. The notice was published

in the Herald Palladium on November 6-8, 1999. No public comments were

received.

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

exigent circumstances, and final determination of no significant

hazards consideration are contained in a Safety Evaluation dated

November 15, 1999.

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

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Section Chief: Claudia M. Craig.

Dated at Rockville, Maryland, this 23rd day of November 1999.

For the Nuclear Regulatory Commission.

Suzanne C. Black,

Deputy Director, Division of Licensing Project Management, Office of

Nuclear Reactor Regulation.

[FR Doc. 99-31037 Filed 11-30-99; 8:45 am]

BILLING CODE 7590-01-P