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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

Commission (the Commission or NRC staff) is

[[Page 54371]]

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 September 11, 1999, through September 24,

1999. The last biweekly notice was published on September 22, 1999 (64

FR 51343 ).

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 and

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and should cite the publication date and page number of

this Federal Register notice. Written comments may also be delivered to

Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of

written comments received may be examined at the NRC Public Document

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

filing of requests for a hearing and petitions for leave to intervene

is discussed below.

By November 5, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Maryland

Date of amendments request: September 1, 1999.

Description of amendments request: The proposed amendment requests

the following changes to the Technical Specifications:

1. Change the definition of Azimuthal Power Tilt in Technical

Specification 1.1;

2. Correct the peak linear heat rate safety limit in Technical

Specification 2.1.1.2;

3. Correct the DC voltage range listed in Surveillance Requirements

3.8.3.9 and 3.8.1.15;

4. Correct the loss of voltage and degraded voltage settings in

Surveillance Requirement 3.3.6.2;

5. Correct the list of core operating limits in Technical

Specification 5.6.5.a;

6. Correct a note on Technical Specification Figure 2.1.1-1;

7. Remove references to Unit 2, Cycle 12 in various Technical

Specifications; and

8. Correct a typographical error in Technical Specification 5.6.

Specifically, the Proposed Technical Specifications are as follows:

1. Technical Specification 1.1 is proposed to be changed to replace

the definition of Azimuthal Power Tilt with a new definition.

2. Technical Specification 2.1.1.2 is proposed to be changed by

replacing the peak linear heat rate safety limit with less than or

equal to 22kW/ft.

3. Technical Specification SR 3.3.6.2 is proposed to be changed by

replacing the degraded voltage function with transient degraded voltage

and steady-state degraded voltage functions.

4. Technical Specification SRs 3.8.1.9 and 3.8.1.15 are proposed to

be changed by replacing the steady-state voltage range with the range

of greater than or equal to 4060 volts and less than or equal to 4400

volts.

5. Technical Specification 5.6.5.a is proposed to be changed by

adding Technical Specifications 3.1.4 and 3.3.1 to the list.

6. Technical Specification Figure 2.1.1-1 is proposed to be changed

by removing the reference to Figure B2.1-1.

7. Various Technical Specifications and Figure 2.1.1-1a.

8. Technical specification 5.6.5.b, Item 41.ii is proposed to be

changed by correcting CEN-199(B)-P to CEN-119(b)-P.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

of consequences of an accident previously evaluated.

Change the Definition of Azimuthal Power Tilt

In their Infobulletin 97-07, Revision 1, Asea Brown Boveri,

Inc.,--Combustion Engineering, Inc. (ABB-CE) stated that they had

found a discrepancy in the Technical Specification definition of

azimuthal power tilt. This discrepancy was found to exist in all CE

Nuclear Steam supply System analog plants that use CECOR for

monitoring and surveillance, and that use ABB-CE safety analysis

methodology. Calvert Cliffs is one of those plants.

The value of Tq (Azimuthal tilt magnitude) as used in the

azimuthal power tilt formula now in Technical Specification 1.1 is

not conservative in all cases. With the proposed definition, Tq is

the maximum fractional increase in power that can occur anywhere in

the core because of tilt. Since Tq is the maximum value, it is

consistently conservative. This is the appropriate measured value of

tilt to be used in verifying that the tilt assumed in establishing

safety limits has not been exceeded.

Therefore, changing the definition of azimuthal power tilt as

proposed will not involve a significant increase in the probability

of consequences of an accident previously evaluated.

Correct the Peak Linear Heat Rate Safety Limit

When Improved Standard Technical Specifications (ITS) were

written, the peak linear heat rate safety limit of [less than or

equal to] 21 kW/ft was inadvertently written in Technical

specification 2.1.1.2. the correct number is [less than or equal to]

22kW/ft. the peak linear heat rate safety limit was established at

[less than or equal to] 22 kW/ft in License Amendment Nos. 88 (Unit

1) and 61 (Unit 2). This number was valid for both units at the time

of implementation of ITS.

Therefore, changing the peak linear heat rate safety limit to a

number previously approved by the Nuclear Regulatory Commission

(NRC) will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Correct the Diesel Generator Loss of Voltage and Degraded Voltage

Settings

When the ITS were written, a single set of numbers for the

degraded voltage function was provided in Technical Specification

Surveillance Requirement (SR) 3.3.6.2. The degraded voltage function

should have been expressed as transient degraded voltage and steady-

state degraded voltage. This separation of two types of degraded

voltage functions was approved in License Amendment Nos. 226 (Unit

1) and 200 (Unit 2), which were issued before the ITS were approved.

Therefore, changing the degraded voltage function to the

transient degraded voltage and steady-state degraded voltage

functions previously approved by the NRC will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Correct the Diesel Generator Voltage Range

Technical Specification SRs 3.8.1.9 and 3.8.1.15 require each

diesel to be started from a stand-by condition. Surveillance

requirement 3.8.1.9 requires that the generator reach [greater than

or equal to] 3740 volts within 10 seconds. After steady-state

conditions are reached, both SRs require the generator to maintain a

voltage range of greater than 3740 volts and [less than or equal to]

4580 volts.

[[Page 54373]]

The Baltimore Gas and Electric Company ITS conversion added

voltage requirements to SRs 3.8.1.9 and 3.8.1.15 consistent with SR

3.8.1.3. License Amendment Nos. 226 and 200 changed the voltage

requirement for SR 3.8.1.3 to [greater than or equal to] 4060 volts

and [less than or equal to] 4400 volts. The voltage was not

corrected in SRs 3.8.1.9 and 3.8.1.15 when the Technical

Specifications were changed to ITS.

Therefore, changing the voltage in SRs 3.8.1.9 and 3.8.1.15 to

voltage previously approved by the NRC will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Correct the List of Core Operating Limits

Technical Specification 5.6.5.a lists Technical Specifications

that are to be included in the core operating limits and documented

in the Core Operating Limits Report (COLR). In the transition to

ITS, Technical Specifications 3.1.4 (Control Element Assembly

Alignment) and 3.3.1 (Reactor Protective System--Operating) were

inadvertently omitted from the list. The complete list is currently

in the COLR.

Therefore, restoring Technical Specification 5.6.5.a to a list

previously approved by the NRC will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Correct Figure 2.1.1-1

A note of Technical Specification Figure 2.1.1-1 was changed in

License Amendment Nos. 227 (Unit 1) and 201 (Unit 2) (ITS) to delete

reference to Figure B2.1-1. Figure B2.1-1 was deleted from the

Technical Specification Bases in the transition to ITS. In License

Amendment Nos. 228 (Unit 1) and 202 (Unit 2), an old version of

Figure 2.1.1-1 was used, and the reference to Figure B2.1-1 was thus

inadvertently put back in the note. The proposed correction will

replace the reference to Figure B2.1-1 with the wording approved in

License Amendment Nos. 227 and 201.

Therefore, returning the note in Figure 2.1.1-1 to the wording

previously approved by the NRC will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Remove References to Unit 2, Cycle 12

License Amendment Nos. 228 and 202 added notes to indicate areas

in the Technical Specifications that had special application to

Cycle 12 of Unit 2 only. Cycle 12 of Unit 2 ended in May 1999. Since

these notes no longer have application, they are proposed to be

removed. Additionally, Figure 2.1.1-la applies only to Unit 2, Cycle

12, and it is proposed to be removed.

Therefore, removal of information no longer applicable to either

unit is an administrative change and will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Correct a Typographical Error

Technical Specification 5.6.5.b, Item 41.ii is being corrected

to change the number of the publication ``BASSS, Use of the Incore

Detector System to Monitor the DNB-LCO on Calvert Cliffs Unit 1 and

Unit 2'' from CEN-199(B) to CEN-119(B)-P. Correction of a

typographical error does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

of accident from an accident previously evaluated.

Change the Definition of Azimuthal Power Tilt

In their Infobulletin 97-07, Revision 1, ABB-CE stated that they

had found a discrepancy in the Technical specification definition of

azimuthal power tilt. This discrepancy was found to exist in all CE

Nuclear Steam Supply System analog plants that use CECOR for

monitoring and surveillance and that use ABB-CE safety analysis

methodology. Calvert Cliffs is one of those plants.

The value of Tq (azimuthal tilt magnitude) as used in the

azimuthal power tilt formula now in Technical specification 1.1 is

not always the most conservative in all cases. With the proposed

definition, Tq is the maximum fractional increase in power that can

occur anywhere in the core because of tilt. Since Tq is the maximum

value, it is conservative. This is the appropriate measured value of

tilt to be used in verifying that the tilt assumed by ABB-CE in

establishing safety limits has not been exceeded.

Therefore, changing the definition of azimuthal power tilt as

proposed will not create the possibility of a new or different type

of accident from any accident previously evaluated.

Correct the Peak Linear Heat Rate

When the ITS were written, a value of peak linear heat rate

[less than or equal to] 21 kW/ft was inadvertently written in

Technical Specification 2.1.1.2. The correct number is [less than or

equal to] 22 kW/ft. The required peak linear heat rate was

established at [less than or equal to] 22 kW/ft in License Amendment

Nos. 88 and 61. This number was valid for both units at the time of

implementation of ITS.

Therefore, changing the value of peak linear heat rate to a

value previously approved by the NRC will not create the possibility

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

evaluated.

Correct the Diesel Generator Loss of Voltage and Degraded Voltage

Settings

When the ITS were written, a single set numbers for the degraded

voltage function was provided in Technical specification SR 3.3.6.2.

The degraded voltage function should have been expressed as

transient degraded voltage and steady-state degraded voltage. This

separation of two types of degraded voltage functions was approved

in License Amendment Nos. 226 and 200, which were issued before the

ITS were approved.

Therefore, changing the degraded voltage function to the

transient degraded voltage and steady-state degraded voltage

functions previously approved by the NRC will not create the

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

previously evaluated.

Correct the Diesel Generator Voltage Range

Technical Specification SRs 3.8.1.9 and 3.8.1.15 require that

each diesel be started from a stand-by condition. Surveillance

Requirement 3.8.1.9 requires that the generator reach [greater than

or equal to] 3740 volts within 10 seconds. After steady-state

conditions are reached, both SRs require the generator to maintain a

voltage range of greater than 3740 volts and [less than or equal to]

4580 volts.

The Baltimore Gas and Electric Company ITS conversion added

voltage requirements to SRs 3.8.1.9 and 3.8.1.15 consistent with SR

3.8.1.3. License Amendment Nos. 226 and 200 changed the voltage

requirement for SR 3.8.1.3 to [greater than or equal to] 4060 volts

and [less than or equal to] 4400 volts. The voltage was not

corrected in SRs 3.8.1.9 and 3.8.1.15 when the Technical

Specifications were changed to ITS.

Therefore, changing the voltage in SRs 3.8.1.9 and 3.8.1.15 to a

voltage previously approved by the NRC will not create the

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

previously evaluated.

Correct the List of Core Operating Limits

Technical Specification 5.6.5.a lists Technical specifications

that are to be included in the core operating limits and documented

in the COLR. In the transition to ITS, Technical Specifications

3.1.4 (Control Element Assembly Alignment) and 3.3.1 (Reaction

Protective System--Operating) were inadvertently omitted from the

list. The complete list is currently in the COLR.

Therefore, restoring Technical Specification 5.6.5.a to a list

previously approved by the NRC will not create the possibility of a

new or different type of accident from any accident previously

evaluated.

Correct Figure 2.1.1-1

A note on Technical Specification Figure 2.1.1-1 was changed in

License Amendment Nos. 227 and 201 (ITS) to delete reference to

Figure B2.1-1. Figure B2.1-1 was deleted from the Technical

Specification Bases in the transition of ITS. In License Amendment

Nos. 228 and 202, an old version of Figure 2.1.1-1 was used, and the

reference to Figure B2.1-1 was thus inadvertently put back in the

note. The proposed correction will replace the reference to Figure

B2.1-1 with the wording approved in License Amendment Nos. 227 and

201.

Therefore, removal of information no longer applicable to either

unit is an administrative change and will not create the possibility

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

evaluated.

Remove References to Unit 2, Cycle 12

License Amendment Nos. 228 and 202 added notes to indicate areas

in the Technical Specifications that had special application to

Cycle 12 of Unit 2 only. Cycle 12 of Unit 2 ended in May 1999. Since

these notes no longer have application, they are

[[Page 54374]]

proposed to be removed. Additionally, Figure 2.1.1-1a applies only

to Unit 2, Cycle 12, and is proposed to be removed.

Therefore, removal of information no longer applicable to either

unit is an administrative change and will not create the possibility

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

evaluated.

Correct a Typographical Error

Technical Specification 5.6.5.b, Item 41.ii is being corrected

to change the number of the publication ``BASSS, Use of the Incore

Detector System to Monitor the DNB-LCO on Calvert Cliffs Unit 1 and

Unit 2'' from CEN-199(B)-P to CEN-119(B)-P. Correction of a

typographical error will not create the possibility of a new or

different type of accident from any accident previously evaluated.

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

safety.

Change the Definition of Azimuthal Power Tilt

The margin of safety in this case is whether the azimuthal power

tilt calculation shows the highest (most conservative) value for Tq

(azimuthal tilt magnitude).

The value of Tq as used in the azimuthal power tilt formula now

in Technical Specification 1.1 is not always the most conservative

in all cases. With the proposed definition, Tq is the maximum

fractional increase in power that can occur anywhere in the core

because of tilt. Since Tq is the maximum value, it is conservative.

This is the appropriate measured value of tilt to be used in

verifying that the tilt assumed in establishing safety limits has

not been exceeded.

Therefore, changing the definition of azimuthal power tilt as

proposed will not involve a significant reduction in the margin of

safety.

Correct the Peak Linear Heat Rate Safety Limit

The margin of safety in this case was previously approved by the

NRC in License Amendment Nos. 88 and 61.

Correct the Diesel Generator Loss of Voltage and Degraded Voltage

Settings

The margin of safety in this case was previously approved by the

NRC in License Amendment Nos. 226 and 200.

Correct the Diesel Generator Voltage Range

The margin of safety in this case was previously approved by the

NRC in License Amendment Nos. 226 and 200.

Correct the List of Core Operating Limits

Technical Specification 5.6.5.a lists Technical specifications

that are to be included in the core operating limits and documented

in the COLR. In the transition to ITS, Technical Specifications

3.1.4 (Control Element Assembly Alignment) and 3.3.1 (Reactor

Protective System--Operating) were inadvertently omitted from the

list. The complete list is currently in the COLR.

Therefore, restoring Technical Specification 5.6.5.a to a list

previously approved by the NRC will not involve a significant

reduction in the margin of safety.

Correct Figure 2.1.1-1

A note on Technical Specification Figure 2.1.1-1 was changed in

License Amendment Nos. 227 and 201 (ITS) to delete reference to

Figure B2.1-1. Figure B2.1-1 was deleted from the Technical

Specification Bases in the transition to ITS. In License Amendment

Nos. 228 and 202, an old version of figure 2.1.1-1 was used, and the

reference to Figure B2.1-1 was thus inadvertently put back in the

note. The proposed correction will replace the reference to Figure

B2.1-1 with the wording approved in License Amendment Nos. 227 and

201.

Therefore, returning the note in Figure 2.1.1-1 to the wording

previously approved by the NRC will not involve a significant

reduction in the margin of safety.

Remove References to Unit 2, Cycle 12

License Amendment Nos. 228 and 202 added notes to indicate areas

in the Technical Specifications that had special application to

Cycle 12 of Unit 2 only. Cycle 12 of Unit 2 ended in May 1999. Since

these notes no longer have application, they are proposed to be

removed. Additionally, Figure 2.1.1-1a applies only to Unit 2, Cycle

12, and it is proposed to be removed.

Therefore, removal of information no longer applicable to either

unit is an administrative change and will not involve a significant

reduction in the margin of safety.

Correct a Typographical Error

Technical specification 5.6.5.b, Item 41.ii is being corrected

to change the number of the publication ``BASSS, Use of the Incore

Detector system to Monitor the DNB-LCO on Calvert cliffs Unit 1 and

Unit 2'' from CEN-199(B)-P to CEN-119(B)-P. Correction of a

typographical error will 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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Section Chief: S. Singh Bajwa.

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

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

Carolina

Date of amendment request: August 26, 1999.

Description of amendment request: The proposed amendment would

revise TS 3/4.9.4, ``Containment Building Penetrations,'' and its

associated Bases to allow penetrations which provide direct access from

the containment atmosphere to the outside atmosphere to remain open

during refueling operations provided certain administrative controls

are met.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Containment is not an accident initiating system as described in

the Final Safety Analysis Report. This change is applicable only in

Mode 6 during Core Alterations or movement of irradiated fuel (which

occurs when the unit is shutdown). The proposed change will not

modify equipment used for fuel movement or core alterations within

the HNP [Harris Nuclear Plant] Containment Building. Administrative

controls will be used to isolate containment in the event of a fuel

handling accident. The consequences of a Fuel Handling Accident

inside containment will increase as a result of this change.

However, the proposed administrative controls will require closure

of containment prior to exceeding standard review plan dose limits

due to a radiological release from a design basis fuel handling

accident.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change provides for administrative controls and

operating restrictions for air lock doors consistent with previous

guidance authorized by the Commission for similar nuclear power

plants. Containment is not an accident initiating system as

described in the Final Safety Analysis Report. Fuel Handling

Accidents have been previously analyzed for the Harris Nuclear

Plant.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

Administrative controls will be used to isolate containment in

the event of a fuel handling accident. The proposed administrative

controls will require closure of containment prior to exceeding

standard review plan dose limits due to a radiological release from

a design basis fuel handling accident.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

[[Page 54375]]

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

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

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

Raleigh, North Carolina 27602.

NRC Section Chief: Sheri R. Peterson.

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

No. 2, Pope County, Arkansas

Date of amendment request: July 29, 1999.

Description of amendment request: The proposed change to the

Arkansas Nuclear One, Unit 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 a scope expansion criteria

to address unexpected conditions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

An evaluation of the proposed change has been performed in

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

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

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

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

Probability or Consequences of an Accident Previously Evaluated.

This change has no actual impact on any previously analyzed

accident in the final safety analysis report (FSAR). A double-ended

break of one steam generator tube is postulated as part of the ANO-2

design basis accident evaluation. The change permits Entergy

Operations to determine the appropriate scope and expansion criteria

for a special steam generator tube inspection that is being

performed at a frequency more conservative than that of the

augmented inservice inspection program included in the TSs

[Technical Specifications]. The special inspection will find and

repair certain steam generator tubing flaws that would otherwise

remain in service until the next scheduled refueling outage. The

increased inspection frequency reduces the probability that a flaw

in a steam generator tube could grow to a size that would affect the

leakage or structural integrity of the tube. The augmented inservice

inspection program contained in the TSs is not being modified.

This change does not modify any parameter that will increase

radioactivity in the primary system or increase the amount of

radioactive steam released from the secondary safety valves or

atmospheric dump valves in the event of a tube rupture.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

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

Different Kind of Accident from any Previously Evaluated.

The scope of this change does not establish a potential new

accident precursor. The design basis accident analyses for ANO-2

include the consequences of a double-ended break of one steam

generator tube which bounds other postulated failure mechanisms. The

proposed change would permit determination of alternate inspection

criteria for a special inspection which is in addition to the

periodic inservice inspections required by the TSs. The equipment

used in the special inspection would not affect any plant components

differently than those used for current TS required inspections.

Therefore, this change does not create the possibility of a new

or different kind of accident from any previously evaluated.

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

Margin of Safety.

As previously stated, a double-ended rupture of one steam

generator tube is accounted for in the ANO-2 design basis accident

analysis. Considering that the 2P99 special inspection is in

addition to the inservice inspection program defined in the ANO-2

TSs and that leakage detection capability is not being modified,

performance of a special inspection of any scope will increase the

margin of safety over the current TS requirements.

Therefore, this change does not involve a significant reduction

in the margin of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve a significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

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

NRC Section Chief: Robert A. Gramm.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook

Nuclear Plant, Unit 1, Berrien County, Michigan

Date of amendment request: August 17, 1999.

Description of amendment request: The proposed amendment would

remove the voltage-based repair criteria, F* repair criteria, and

sleeving methodologies from the Unit 1 Technical Specifications (T/S)

and clarify the Bases sections accordingly.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This change removes the interim steam generator tube plugging

criteria from the T/S and reinstates the original T/S criteria

consistent with Unit 2 (which does not have significantly degraded

steam generators). The current T/S allow for continued operation

with tubes that demonstrate indications per F* and voltage-based

criteria. The basis used to justify the interim criteria is specific

to the Unit 1 original steam generators (OSGs) and does not apply to

the replacement steam generators (RSGs).

The proposed change returns the plugging criteria for the steam

generator tubes to the original licensing basis. The criteria are in

accordance with NUREG-0452, (old) ``Standard Technical

Specifications.'' The plugging criteria are based on a minimum wall

thickness due to wastage as determined by ASME [American Society of

Mechanical Engineers] Section XI. The proposed change is

conservative in nature because it does not allow for continued

operation with F* and voltage-based degraded tubes. Because of this,

the probability of a steam generator tube rupture (SGTR) is not

increased.

The potential for a SGTR is also not increased as demonstrated

in the qualification analysis and testing for the RSGs. The program

for periodic in-service inspection monitors the integrity of the SG

tubing to provide reasonable assurance that there is sufficient time

to take proper and timely corrective action if any tube degradation

is detected. The tube inspections themselves are not initiators of a

SGTR. Therefore, this change is not expected to increase the

probability of a SGTR during normal or accident conditions.

Unit 1 will continue to apply the T/S maximum primary-to-

secondary leakage limit of 150 gallons per day (gpd) through any one

SG to minimize the potential for excessive leakage. The EPRI

[Electric Power Research Institute]-recommended 150 gpd limit

[[Page 54376]]

provides for leakage detection and plant shutdown in the event of an

unexpected tube leak and minimizes the potential for excessive

leakage or tube burst in the event of main steamline break (MSLB) or

loss-of-coolant accident (LOCA) conditions. This lower limit is more

restrictive than the limit (500 gpd per SG and total leakage of 1440

gpd) utilized for determination of offsite dose and also provides

further assurance that the probability of a SGTR is not increased.

The design basis doses calculated for postulated accidents

involving degradation of SG tubes, such as SGTR and MSLB accidents,

as presented in UFSAR chapter 14 accident analysis, have been

evaluated. The SGTR consequences continue to be bounded by the

design basis analyses due to the allowable leakage rate specified by

this change. The proposed T/S leakage rate is maintained at 150 gpd

per SG. However, the maximum leakage of 500 gpd per SG and total

leakage of 1440 gpd for all four generators was used to determine

offsite dose in UFSAR chapter 14. The MSLB consequences are

decreased by installation of the RSGs due to the reduction in

primary-to-secondary leakage during the MSLB. Under the approved

interim plugging criteria, a leak rate of 8.4 gpm was determined to

be the upper limit for allowable primary-to-secondary leakage in the

faulted steam generator. This leakage, combined with the 150 gpd

leakage from the non-faulted SGs, was determined to limit the

offsite dose to 10% of the 10 CFR 100 limits. Following replacement

of the SGs, the leakage is limited during the MSLB to 150 gpd for

both the faulted and unfaulted SGs. Therefore, the Unit 1 MSLB dose

will be bounded by the current Unit 2 dose analysis, which is less

than 10% of 10 CFR 100 limits.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

Removing application of voltage-based repair criteria, F* repair

criteria, and sleeving methodologies upon installation of the RSGs

will not introduce significant or adverse changes to the plant

design basis that could lead to a new or different kind of accident

being created. This change does not change the overall objective of

surveillance activities--maintaining the structural integrity of

this portion of the reactor coolant system. The surveillance

activities are performed during outages. The proposed change in the

surveillance program returns the program to the initial licensing

basis. No new failures are created.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Removing the application of voltage-based and F* repair criteria

and sleeving methodologies does not involve a reduction in the

margin of safety. The RSG tubing has been shown to retain adequate

structural and leakage integrity during normal, transient, and

postulated accident conditions consistent with GDC 14, 15, 30, 31,

and 32 of 10 CFR [Part] 50 [A]ppendix A. The RSG tubing has been

designed and evaluated consistent with the ASME Section III, 1989

edition. The proposed plugging criteria are based on ASME Section XI

and do not allow for operation with indications identified by F* and

voltage-based criteria. The proposed program for periodic in-service

inspection of the RSGs monitors the integrity of the SG tubing to

provide reasonable assurance that there is sufficient time to take

proper and timely corrective action if any tube degradation is

present. The proposed program is consistent with NUREG-0452 and was

the basis for the original Unit 1 T/S surveillance program.

The proposed change maintains the T/S maximum primary-to-

secondary leakage at 150 gpd per generator to minimize the potential

for excessive leakage. This limit provides for leakage detection and

shutdown in the event of an unexpected tube leak and minimizes the

potential for excessive leakage or tube burst in the event of a MSLB

or LOCA. Because this limit is maintained, the margin of safety is

maintained.

Therefore, it is concluded that this change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Attorney for licensee: Jeremy J. Euto, Esq., 500 Circle Drive,

Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.

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: September 10, 1999.

Description of amendment requests: The proposed amendments would

revise Technical Specification (T/S) 3/4.4.7 so that the surveillance

requirement does not need to be performed when the reactor is defueled

with no forced circulation. The proposed revision to T/S 3/4.4.7 also

includes changes to Tables 3.4-1 and 4.4-3. A change is proposed to

Unit 1 T/S Table 4.4-3 to revise the reactor coolant system (RCS)

chemistry sampling frequency from three times per 7 days with a maximum

interval of 72 hours to a frequency of at least once per 72 hours. An

editorial change to Unit 1 Tables 3.4-1 and 4.4-3 would relocate the

asterisk for the footnote to a position adjacent to the parameter

``dissolved oxygen,'' from its current position next to the allowable

chemistry limit in Table 3.4-1 and the analysis frequency in Table 4.4-

3. An editorial change would also correct the footnote for Table 3.4-1

for Unit 1 and Unit 2 by making the word ``limit'' plural, as it

applies to both the steady-state and transient limits.

Changes are also proposed to revise Surveillance Requirement

4.11.2.2 by deleting the phrase ``by analysis of the Reactor Coolant

System noble gases.''

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 changes to the RCS chemistry sampling requirements

do not affect the probability of a loss-of-coolant accident or steam

generator tube rupture, which are evaluated in Sections 14.3 and

14.2.4, respectively, of the Updated Final Safety Analysis Report

(UFSAR). RCS contaminant limits are maintained to reduce the

potential for RCS leakage or failure due to corrosion. Sampling the

RCS for contaminants does not initiate an accident. Deleting the

requirement to obtain samples when the reactor is defueled does not

modify any plant equipment or affect plant operation and therefore

does not introduce any new accident initiators or precursors.

Suspension of RCS chemistry sampling when the reactor is defueled

does not increase the potential for RCS leakage or failure because

the corrosive effects of the contaminants is minimal during this

low-temperature, low-pressure condition. To ensure elevated

contaminant levels would be detected and corrected prior to

subjecting the system to a high-temperature condition, chemistry

sampling will be reinstated within 72 hours of re-establishing

forced circulation and prior to entering Mode 6. Removing the

restriction for analyzing primary coolant chemical contaminants at

least three times every seven days does not change the maximum

surveillance interval. This change allows the sample to be collected

two or three times per week, consistent with the maximum 72-hour

interval. The 72-hour sampling and analysis interval is consistent

with the current requirement in the Unit 2 T/S, and industry

guidance in NUREG-0452, ``Standard Technical Specifications.'' The

72-hour interval continues to provide adequate assurance that

concentrations in excess of the limits are detected in sufficient

time to take corrective actions. Therefore, the probability of

occurrence of a previously evaluated accident is not increased.

This change does not alter the quantity of radioactive material

in any system during normal plant operation, the amount of

[[Page 54377]]

shielding provided by plant systems, or the mitigative capabilities

of any system following an event. Therefore, the consequences of a

previously evaluated accident are not increased.

The editorial changes to the RCS chemistry T/S provide

consistency between the Unit 1 and Unit 2 T/S and the Standard

Technical Specifications. These changes do not affect the design or

operation of any system, structure, or component in the plant. The

accident analysis assumptions and results are unchanged. No new

failures or interactions are created.

The amount of radioactive material in the gas storage tanks is

controlled to ensure that, in the event of a rupture of one of these

tanks, the resulting total body exposure to an individual at the

nearest site boundary would not exceed 0.5 rem. The accidental waste

gas release event is summarized in Section 14.2.3 of the UFSAR.

Sampling to determine the radioactivity levels in the tanks does not

initiate an accident or identify any accident precursors. The

increased sampling flexibility does not change the method of

operating the waste gas system, nor does it modify any interfaces

with other plant systems. Therefore, this change does not increase

the probability of occurrence of an accidental waste gas release

event.

Implementation of a different sampling method does not change

the maximum quantity of radioactive material specified in the T/S

Limiting Condition for Operation (LCO). The sampling method has no

effect on normal plant gaseous radwaste activities, so the

composition of the radioactive gaseous nuclides present in the tank

at the time of the event is not affected. As the proposed revision

allows a change to the method of sampling but does not affect the

radioactivity limit for the gas storage tanks, the proposed change

does not increase the consequences of an accidental waste gas

release event.

Therefore, the probability of occurrence or the consequences of

accidents previously evaluated are not increased.

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

kind of accident from any accident previously evaluated?

The proposed changes to revise the RCS chemistry sampling

frequency and to suspend RCS chemistry sampling when the reactor is

defueled with no forced circulation does not change the method of

operating any equipment or the operational limits of any equipment.

The proposed changes do not introduce any new failure mechanisms to

the RCS or any other plant systems. The proposed change does not

involve any physical alterations to any plant equipment, and causes

no change in the method by which any plant system performs its

function. Editorial changes to footnotes for Tables 3.4-1 and 4.4-3

provide consistency between the T/S for Unit 1 and Unit 2, but do

not change the methods of operating any equipment or introduce any

new failure mechanisms.

The proposed change to eliminate the prescriptive waste gas tank

sampling method does not introduce any new failure mechanisms to the

waste disposal system, involve any physical changes to the waste

disposal system or any other plant systems, or change the way any

plant systems are operated. This change does not change any

interfaces between the waste disposal system and any other plant

systems. The proposed changes continue to ensure the system is

operated within the existing limit established by the T/S LCO. Thus,

no adverse safety considerations are introduced by this proposed

change to the T/S.

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?

The margin of safety pertinent to the RCS chemistry surveillance

is related to the concentration of chemical contaminants that would

expedite corrosion of the RCS piping and components and the period

of time during which the system is allowed to operate outside the T/

S limits. The proposed changes to the RCS chemistry surveillance do

not alter either of these criteria. These proposed changes do not

affect any safety limits or T/S parameter limits. The proposed

changes do not introduce new equipment, equipment modifications, or

new or different modes of plant operation. These changes do not

affect the operational characteristics of any equipment or systems.

The editorial changes to footnotes for Tables 3.4-1 and 4.4-3

provide consistency between the T/S for Unit 1 and 2, but do not

affect the acceptance criteria or surveillance frequencies for this

T/S.

The margin of safety pertinent to the waste gas storage tanks is

related to the quantity of radioactivity that would be released in

the unlikely event of a tank rupture. The proposed change to the gas

storage tank T/S eliminates the prescriptive sampling methodology,

but does not affect the requirement to periodically quantify the

radioactive gaseous material in the gas storage tanks. The proposed

change does not affect the quantity of radioactivity allowed in the

gas storage tanks, nor does it alter the methodology, assumptions,

or results of any safety analyses. The proposed change to delete the

prescriptive sampling method does not affect any safety limits or T/

S parameter limits.

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.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Attorney for licensee: Jeremy J. Euto, Esq., 500 Circle Drive,

Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.

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

NASA Test Reactor, Erie County, Ohio

Date of amendment request: March 25, 1999, as supplemented by

letter dated August 10, 1999.

Description of amendment request: The proposed amendment would

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

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed amendment will change the name of the Licensee for

the Plum Brook Reactor Facility (PBRF) TR-3 license, a possession

only license, from Lewis Research Center (LeRC) to the Glenn

Research Center (GRC). The amendment request is necessary because

NASA has changed the name of the Lewis Research Center to the Glenn

Research Center at Lewis Field under legislative action and signed

into law (sec. 434, P.L. 105-276, 112 Stat. 2461) on October 21,

1998. The effective date of this name change was March 1, 1999.

NASA, GRC will retain the PBRF license and the responsibility to

continue maintaining the PBRF Reactor Facility in a safe protected

storage mode under the current TR-3 possess-but-not-operate license.

In addition, the current plans to provide a PBRF decommissioning

plan to the NRC by the end of CY 1999 and the eventual

decommissioning by the end of CY 2007 have not changed.

There will be no change in the funding status of the GRC in

either maintaining the PBRF facility in the safe protected storage

mode or the eventual decommissioning. NASA, as a government agency,

remains responsible for the continuing funding of both activities.

In addition, there will be no change in the personnel who are

responsible for maintaining the present TR-3 license or in

developing the PBRF Decommissioning Plan.

The proposed amendment does not require any physical change to

the PBRF Facility, changes to the Technical Specifications or

procedures under the PBRF TR-3 License other than the name change

from LeRC to GRC. The proposed change does not increase the

probability of any accident or increased risk to the public safety.

Therefore, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident condition

previously evaluated.

(2) Would not create the possibility of a new or different kind

of accident from any accident previously evaluated.

The proposed amendment does not modify the PBRF facility

configuration or licensed activities. Therefore, no additional

accident conditions are introduced.

[[Page 54378]]

Therefore, the proposed amendment does not involve a significant

increase in the probability or consequence of an accident.

(3) Would not involve a significant reduction in a margin of

safety.

This amendment is required because of the name change from LeRC

to GRC. NASA will continue to be financially responsible to maintain

the PBRF Facility under the existing TR-3 License.

Furthermore, the GRC personnel for the eventual PBRF

decommissioning and contract support personnel reporting to GRC will

continue to be technically qualified to maintain the PBRF under the

safe protected storage mode. There has been no effective change in

the personnel who will be responsible to implement the eventual

decommissioning effort that will be required under the future PBRF

Decommissioning Plan.

Plum Brook's existing qualified contractors remained in place

following the name change. The requested amendment does not involve

any changes in the performance of current licensed activities and

these activities will continue in their current form without changes

or interruptions of any kind.

The proposed amendment does not alter any margin of safety

because it does not involve any changes in the PBRF Facility or

licensed activities under the TR-3 License. All activities will

continue in the current form without changes or interruptions of any

kind as a result of the name.

Therefore, the proposed amendment does not involve a significant

reduction in a margin of safety.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: N/A.

Attorney for licensee: Elias T. Naffah, MS 500-118, NASA, Glenn

Research Center, 21000 Brookpark Road, Cleveland Ohio 44135.

NRC Branch Chief: Ledyard B. Marsh.

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

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

Date of amendment request: July 16, 1999.

Description of amendment request: Proposed relocation of Technical

Specifications 3/4.9.3.2, ``Refueling Operations, Spent Fuel

Temperature,'' 3/4.9.3.3, ``Refueling Operations, Decay Time,'' 3/

4.9.5, ``Refueling Operation, Communications,'' 3/4.9.6, ``Refueling

Operation, Crane Operability--Containment Building,'' and 3/4.9.7,

``Refueling Operations, Crane Travel--Spent Fuel Storage Building,'' to

the Millstone, Unit No. 2 Technical Requirements Manual. The associated

Bases pages and index pages will be modified to address the 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. The NRC staff has reviewed the licensee's analysis

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

presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Technical Specification 3/4.9.3.2, ``Refueling Operations, Spent

Fuel Pool Temperature,'' is proposed to be relocated to the TRM

where future changes will be controlled in accordance with 10 CFR

50.59. This specification limits spent fuel pool temperature to be

less than or equal 140 deg.F to ensure the resin in the spent fuel

cooling demineralizers will not degrade and the temperature and

humidity are compatible with personnel comfort and safety

requirements. Additionally, the requirement ensures that the design

temperature of the fuel pool cooling system, liner/building

structures, and racks is not exceeded. Relocation of this Technical

Specification to the TRM does not imply any reduction in its

importance in limiting the spent fuel pool bulk temperature to be

less than or equal to 140 deg.F. Spent fuel pool bulk temperature

is a design bases process variable which is used to establish the

required heat removal capabilities of the spent fuel heat removal

system. In the unlikely event of total loss of cooling water flow to

the spent fuel pool, the pool water temperature may reach 212 deg.F

within approximately 9 hours and will result in a boiling condition.

This event does not represent a challenge to the fuel cladding, as a

fission product barrier, unless the fuel becomes uncovered. The

requirement on storage pool water level is covered by Technical

Specification 3/4.9.12, ``Storage Pool Water Level,'' which requires

a minimum of 23 feet of water over the top of irradiated fuel

assemblies. Therefore, spent fuel pool bulk temperature is not by

itself a process variable that is an initial condition of a design

basis accident. This Technical Specification does not cover a

process variable, design feature, or operating restriction that is

an initial condition of a design basis accident or transient

analysis that either assumes the failure of or presents a challenge

to the integrity of a fission product barrier. It does not cover a

structure, system, or component that is part of the primary success

path which functions or actuates to mitigate a design basis accident

or transient that either assumes the failure of or presents a

challenge to the integrity of a fission product barrier. The

proposed change will not alter the way pool temperature is measured,

nor will it alter any of the assumptions used in the spent fuel pool

fuel handling accident analysis. Relocation of this Technical

Specification to the TRM does not degrade the performance of any

safety systems or prevent actions assumed in the accident analysis,

nor does it alter any of the assumptions made in the analysis that

could increase the consequences of accidents. Therefore, this change

will not significantly increase the probability or consequences of

an accident previously evaluated.

Technical Specification 3/4.9.3.3, ``Refueling Operations, Decay

Time,'' is proposed to be relocated to the TRM where future changes

will be controlled in accordance with 10 CFR 50.59. This

specification requires the reactor to remain in Mode 5 or 6 until

the most recent core offload has decayed a sufficient time to ensure

alternate cooling is available during this time to cool the spent

fuel pool should a failure occur in the Spent Fuel Pool Cooling

System. Alternate cooling would be provided by the Shutdown Cooling

System. Relocation of this Technical Specification to the TRM does

not imply any reduction in its importance in insuring that the most

recent core offload has decayed a sufficient time. If the

requirement to remain in Mode 5 or 6 until the most recent core

offload has decayed for 504 hours is not satisfied, the spent fuel

pool cooling system may not have the capability to remove decay heat

and stay below the Technical Specification limit of 140 deg.F. In

the unlikely event of total loss of cooling water flow to the spent

fuel pool, the pool water temperature may reach 212 deg.F in less

than 9 hours and will result in a boiling condition. This event does

not represent a challenge to the fuel cladding, as a fission product

barrier, unless the fuel becomes uncovered. The requirements on

storage pool water level is covered by Technical Specification 3/

4.9.12, ``Storage Pool Water Level,'' which requires a minimum of 23

feet of water over the top of irradiated fuel assemblies. Therefore,

this requirement to remain in Mode 5 or 6 until the most recent core

offload has decayed for 504 hours is not by itself a process

variable that is an initial condition of a design basis accident.

This Technical Specification does not cover a process variable,

design feature, or operating restriction that is an initial

condition of a design basis accident or transient analysis that

either assumes the failure of or presents a challenge to the

integrity of a fission product barrier. It does not cover a

structure, system, or component that is part of the primary success

path which functions or actuates to mitigate a design basis accident

or transient that either assumes the failure of or presents a

challenge to the integrity of a fission product barrier. The

proposed change will not alter the requirement that the most recent

core offload has decayed a sufficient time, nor will it alter any of

the assumptions used in the spent fuel pool fuel handling accident

analysis. Relocation of this Technical Specification to the TRM does

not degrade the performance of any safety systems or prevent actions

assumed in the accident analysis, nor does it alter any of the

assumptions made in the analysis that could increase the

consequences of accidents. Therefore, this change will not

significantly increase the probability or consequences of an

accident previously evaluated.

Technical Specification 3/4.9.5, ``Refueling Operations,

Communications,'' is proposed to be relocated to the TRM where

future

[[Page 54379]]

changes will be controlled in accordance with 10 CFR 50.59. This

specification requires communication between the control room and

the refueling station, to ensure any abnormal change in the facility

status, as indicated on the control room instrumentation, can be

communicated to the refueling station personnel. Relocation of this

Technical Specification to the TRM does not imply any reduction in

its importance in insuring communication between the control room

and the refueling station. This Technical Specification does not

cover a process variable, design feature, or operating restriction

that is an initial condition of a design basis accident or transient

analysis that either assumes the failure of or presents a challenge

to the integrity of a fission product barrier. It does not cover a

structure, system, or component that is part of the primary success

path which functions or actuates to mitigate a design basis accident

or transient that either assumes the failure of or presents a

challenge to the integrity of a fission product barrier. The

proposed change will not alter the requirement on communication

between the control room and the refueling station, nor will it

alter any of the assumptions used in the spent fuel pool fuel

handling accident analysis. Relocation of this Technical

Specification to the TRM does not degrade the performance of any

safety systems or prevent actions assumed in the accident analysis,

nor does it alter any of the assumptions made in the analysis that

could increase the consequences of accidents. Therefore, this change

will not significantly increase the probability or consequences of

an accident previously evaluated.

Technical Specification 3/4.9.6, ``Refueling Operations, Crane

Operability--Containment Building,'' is proposed to be relocated to

the TRM where future changes will be controlled in accordance with

10 CFR 50.59. This specification ensures the lifting device on the

refueling machine has adequate capacity to lift the weight of a fuel

assembly and a control element assembly, and that an automatic load

limiting device is available to prevent damage to the fuel assembly

during fuel movement. Relocation of this Technical Specification to

the TRM does not imply any reduction in its importance in insuring

that the lifting device on the refueling machine has adequate

capacity. The automatic load limiting device and/or physical stops

are not monitored and controlled during operation, nor are they

assumed to function to mitigate the consequences of a design basis

accident. The automatic load limiting device is checked on a

periodic basis to ensure operability. This Technical Specification,

which ensures the lifting device on the refueling machine has

adequate capacity, does not cover a process variable, design

feature, or operating restriction that is an initial condition of a

design basis accident or transient analysis that either assumes the

failure of or presents a challenge to the integrity of a fission

product barrier. The proposed change will not alter the requirement

that the lifting device on the refueling machine has adequate

capacity, nor will it alter any of the assumptions used in the

accident analysis. Relocation of this Technical Specification to the

TRM does not degrade the performance of any safety systems or

prevent actions assumed in the accident analysis, nor does it alter

any of the assumptions made in the analysis that could increase the

consequences of accidents. Therefore, this change will not

significantly increase the probability or consequences of an

accident previously evaluated.

Technical Specification 3/4.9.7, ``Refueling Operations, Crane

Travel--Spent Fuel Storage Pool Building,'' is proposed to be

relocated to the TRM where future changes will be controlled in

accordance with 10 CFR 50.59. This specification ensures loads in

excess of one fuel assembly containing a control element assembly,

plus the weight of the fuel handling tool, will not be moved over

other fuel assemblies in the spent fuel storage racks. Therefore, in

the event of a drop of this load, the activity released is limited

to that contained in one fuel assembly. Relocation of this Technical

Specification to the TRM does not imply any reduction in its

importance in insuring that loads in excess of 1800 pounds (except

of a consolidated fuel storage box) are prohibited from travel over

irradiated fuel. While this Technical Specification does address an

operating restriction assumed in the accident analysis, there is no

process variable that can be monitored during power operation of the

plant. Crane interlocks and/or physical stops are used to assure

that this requirement is met, but indication of the operation of the

interlocks and/or physical stops is not available in the control

room. These features inhibit movement of the crane so that

monitoring is not necessary. This Technical Specification does not

cover a structure, system, or component that is part of the primary

success path which functions or actuates to mitigate a design basis

accident or transient that either assumes the failure of or presents

a challenge to the integrity of a fission product barrier. The

proposed change will not alter the requirement that the crane

interlocks and/or physical stops are OPERABLE, nor will it alter any

of the assumptions used in the spent fuel pool fuel handling

accident analysis. Relocation of this Technical Specification to the

TRM does not degrade the performance of any safety systems or

prevent actions assumed in the accident analysis, nor does it alter

any of the assumptions made in the analysis that could increase the

consequences of accidents. Therefore, this change will not

significantly increase the probability or consequences of an

accident previously evaluated.

Revision of Index Pages IX and XIII and the proposed change to

Bases sections, by relocating them to the TRM, are administrative

changes. Therefore, this change will not significantly increase the

probability or consequences of an accident previously evaluated. The

proposed changes do not alter how any structure, system, or

component functions. There will be no effect on equipment important

to safety. The proposed changes have no effect on any of the design

basis accidents previously evaluated. Therefore, this License

Amendment Request does not impact the probability of an accident

previously evaluated, nor does it involve a significant increase in

the consequences of an accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes do not alter the plant configuration (no

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

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

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

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

introduce any new failure modes. Therefore, the proposed changes

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

accident from any accident previously evaluated.

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

The proposed relocation of Technical Specification 3/4.9.3.2,

``Refueling Operations, Spent Fuel Pool Temperature,'' to the TRM

does not imply any reduction in its importance in limiting the spent

fuel pool bulk temperature to less than or equal to 140 deg.F. The

proposed change will not alter the way pool temperature is measured.

It will not alter any of the assumptions used in the spent fuel pool

fuel handling accident analysis, nor will it cause any safety system

parameters to exceed their acceptance limit. The proposed relocation

of Technical Specification 3/4.9.3.3, ``Refueling Operations, Decay

Time,'' to the TRM does not imply any reduction in its importance in

insuring that the most recent core offload has decayed a sufficient

time. The proposed change will not alter the requirement that the

most recent core offload has decayed a sufficient time, it will not

alter any of the assumptions used in the spent fuel pool fuel

handling accident analysis, nor will it cause any safety system

parameters to exceed their acceptance limit. The relocation of

Technical Specification 3/4.9.5, ``Refueling Operations,

Communications,'' to the TRM does not imply any reduction in its

importance in insuring communication between the control room and

the refueling station. The proposed change will not alter the

requirement on communication between the control room and the

refueling station, it will not alter any of the assumptions used in

the spent fuel pool fuel handling accident analysis, nor will it

cause any safety system parameters to exceed their acceptance limit.

The relocation of Technical Specification 3/4.9.6, ``Refueling

Operations, Crane Operability--Containment Building,'' to the TRM

does not imply any reduction in its importance in insuring that the

lifting device on the refueling machine has adequate capacity. The

proposed change will not alter the requirement that the lifting

device on the refueling machine has adequate capacity, it will not

alter any of the assumptions used in the accident analysis, nor will

it cause any safety system parameters to exceed their acceptance

limit. The relocation of Technical Specification 3/4.9.7,

``Refueling Operations, Crane Travel--Spent Fuel Storage Pool

Building,'' to the TRM does not imply any reduction in its

importance in insuring that loads in excess of 1800 pounds (except

of a consolidated fuel storage box) are prohibited from travel over

irradiated fuel. The proposed change will not

[[Page 54380]]

alter the requirement that the crane interlocks and/or physical

stops are OPERABLE, it will not alter any of the assumptions used in

the spent fuel pool fuel handling accident analysis, nor will it

cause any safety system parameters to exceed their acceptance limit.

Revision of Index Pages IX and XIII and the proposed change to Bases

sections by eliminating the sections corresponding to the relocated

Technical Specifications are administrative changes. These changes

will not alter any of the assumptions used in the spent fuel pool

fuel handling accident analysis, nor will it cause any safety system

parameters to exceed their acceptance limit. The proposed changes do

not affect any of the assumptions used in the accident analysis, nor

do they affect any operability requirements for equipment important

to plant safety. Therefore, the proposed changes will not result in

a significant reduction in a margin of safety.

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

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Section Chief: James W. Clifford.

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

Unit 1, Montgomery County, Pennsylvania

Date of amendment request: June 7, 1999.

Description of amendment request: The proposed change to the

Technical Specifications (TSs), if approved, will reflect the permanent

deactivated configuration of the ``wet'' instrument reference leg

isolation valve HV-61-102 which originally connected the Drywell Floor

and Equipment Drain Tanks to level instruments outside the containment.

The TS changes affecting TS Table 3.6.3-1, ``Primary Containment

Isolation Valves,'' and its associated notations will reflect the

current plant configuration. More specifically, TS Section 3/4.6.3,

``Primary Containment Isolation Valves,'' Table 3.6.3-1, Penetration

Number 230B will be revised to designate the function of valve HV-61-

102 as ``Deactivated,'' the maximum isolation time for valve HV-61-102

will be eliminated, and notations 1, 23, and 29 will be replaced with a

new notation indicating the permanent configuration of the subject

valve.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below:

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

in the probability or consequences of an accident previously

evaluated.

The closed valve, HV-61-102, has no effect on the function of

the Drywell Sump/Equipment Drain Tanks, other safety-related

systems, or other containment penetrations. The current status of

the valve is locked closed, de-energized, and the motor operator

cannot be accidentally actuated. In addition, the line is capped

downstream of the isolation valve. As described above, the valve is

considered to be in a passive configuration, where a malfunction is

not expected and cannot cause an increase in the probability of a

malfunction to itself or other safety-related equipment. The

potential for increased releases outside the containment due to

breaching of the valve assembly is no greater than that of the

isolation design previously evaluated.

Therefore, the proposed change to the TSs does not involve a

significant increase in the probability of occurrence or

consequences of an accident previously evaluated in the Safety

Analysis Report.

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

new or different kind of accident from any accident previously

evaluated.

The abandoned isolation valve conforms to approved isolation

configurations, and its structural integrity has not been degraded

by the modified configuration. The original function of valve HV-61-

102 was only to provide isolation of the instrument line. Following

the modification, the valve is independent of the function of the

Drywell Sump/ Equipment Drain Tanks, other safety-related systems,

and other penetrations. Since the valve is passive and has no

requirements to be operated, it cannot create a different type of

malfunction on itself or other safety-related systems. In addition,

the valve is specifically designed to isolate and is essentially

passive during accident conditions, it has no activity that could be

the initiator of an accident of a different type.

Therefore, the proposed changes to the TSs do 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 a margin of safety. Isolation valve HV-61-102 in its

proposed permanent configuration meets the margin of safety

described in TS Bases 3/4.6.3 since it is kept closed under all

operational conditions and will not be under the constraint of TS

closing times in order to maintain releases within specifications.

The proposed changes have no impact on any safety analysis

assumptions.

Therefore, the proposed TS changes do not involve a significant

reduction in the margin of safety.

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

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

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.

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

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

Jersey

Date of amendment request: July 23, 1999, as supplemented on

September 13, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification Surveillance Requirement 4.8.1.1.2 to

allow the 24-hour emergency diesel generator endurance run to be

performed during power operation (i.e., Modes 1 and 2) instead of

restricting the test to when the reactor was shutdown.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change to Technical Specification Surveillance

Requirement (SR) 4.8.1.1.2.d.7 (24-hour emergency diesel generator

(EDG) endurance run test) to eliminate the restriction to perform

the test during shutdown conditions does not involve a significant

increase in the probability of any previously evaluated accident.

Although paralleling or connecting the EDG to off-site power for the

test could induce an electrical distribution system perturbation,

the same possibility exists when the EDG is tested during the

monthly 1-hour loaded surveillance test (SR 4.8.1.1.2 a 2). This

risk during testing the EDG monthly at power was reviewed and found

acceptable by the NRC. Further, none of the automatic actuations and

interlocks in the tested portion of the electrical system or the EDG

control system are disabled during the 24-hour endurance run. Thus,

the onsite safety-related electrical system remains protected from

potential faults and perturbations.

The ability and capability [o]f the EDG to perform their safety

function (mitigate the consequences of a previously evaluated

[[Page 54381]]

accident) is also unaffected. This capability was demonstrated not

only by the tests conducted in the EDG manufacturer's plant, but

continue to be demonstrated by surveillance testing performed at the

station.

This testing verifies specific design criteria, which assure

continued EDG operability even during testing. Examples of presently

performed Technical Specification testing that demonstrate the

ability and capability of the EDG to perform its safety functions

are:

SR 4.8.1.1.2. d. 2 requires, in part, that on a load

rejection of greater than 820 KW, the voltage and frequency be

restored to acceptable values within 4 seconds.

This surveillance demonstrates the ability of the EDGs

to withstand a loss of load, as it would occur in a normal

safeguards equipment controller (SEC) actuation, without

compromising its ability to be ready to accept a new loading

sequence and carry its design safety function.

SR 4.8.1.1.2. d. 9 requires, in part, that with the EDG

operating in a test mode (connected to its bus), a simulated safety

injection signal overrides the test mode by (1) returning the diesel

generator to standby operation and (2) automatically energizing the

emergency loads with offsite power.

This surveillance demonstrates the ability of the EDGs to be

disconnected from the grid, if in a test mode, on an accident

signal, and be ready to accept a new loading sequence and carry its

design safety function.

SR 4.8.1.1.2. a. 2 requires, in part, that every 31

days each EDG be demonstrated OPERABLE by synchronizing it to the

grid for greater than or equal to 60 minutes.

Note that this proposed amendment request eliminates a

discrepancy between the current requirement to perform the 24 hour

run during shutdown and SR 4.8.1.1.2.a.2, which would allow a 24

hour run at power.

Additionally, PSE&G performed an assessment of the potentially

added risk of an additional 24 hours of on-line EDG testing. The

unavailability of all three EDGs was increased in the Probabilistic

Safety Analyses (PSA) for both Salem Units 1 and 2 to correspond to

an additional 24 hours per cycle out-of-service time each 18-month

operating cycle. The unavailability was changed from 1.86E-02/year

to 2.0E-2/year. The increase in the baseline internal events core

damage frequency (CDF) was determined to be 1.6E-07 events/year for

both Salem Units 1 and 2. Based on the definition provided in

Regulatory Guide 1.174, Paragraph 2.2.4, this increase is considered

a very small increase in risk (less than 1.0E-06 events/year).

Therefore, the proposed amendment, including proposed

administrative controls, does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed amendment to Technical Specification Surveillance

Requirement 4.8.1.1.2.d.7 (24-hour endurance run test) to eliminate

the restriction to perform the test during shutdown conditions does

not physically modify the facility, introduce a new failure mode, or

propose a different operational mode of the AC electrical power

sources, or Emergency Diesel Generators.

Therefore, the proposed amendment will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The AC Electrical distribution system has been designed to

provide sufficient redundancy and reliability to ensure the

availability of the EDGs to provide the required safety function

under design basis events to protect the power plant, the public and

plant personnel. Specifically, the ability of the EDGs to separate

from the off-site power source has been designed and tested per

Technical Specifications requirements.

Performance of the 24-hour endurance run during power operations

will not affect the availability of any of the required power

sources, nor the capability of the EDGs to perform their intended

safety function. Furthermore, performing the test when the

undervoltage protection of the 4160-V vital buses required by the

Salem Station Technical Specification 3.3.2.1 is operable, provides

for an added level of protection to the EDG that is not available

while shutdown.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

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

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

Date of application for amendments: August 30, 1999 (TS 99-08).

Brief description of amendments: The proposed amendments would

change the Sequoyah (SQN) Technical Specification (TS) requirements to

provide alternatives to the requirement of actually measuring response

times.

Basis for proposed no significant hazards consideration

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

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

of no significant hazards consideration, which is presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

This change to the TS does not result in a condition where the

design, material, and construction standards that were applicable

prior to the change are altered. The same RTS [Reactor Trip System]

and engineered safety feature actuation system (ESFAS)

instrumentation is being used, the time response allocations/

modeling assumptions in the [Final Safety Analysis Report] Chapter

15 analyses are still the same, only the method of verifying time

response is changed. The proposed change will not modify any system

interface and could not increase the likelihood of an accident since

these events are independent of this change. The proposed activity

will not change, degrade or prevent actions, or alter any

assumptions previously made in evaluating the radiological

consequences of an accident described in the Final Safety Analysis

Report. Therefore, the proposed amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

This change does not alter the performance of pressure [or]

differential pressure transmitters, solid state protection system

racks, nuclear instrumentation, or input and output master/slave

relays used in the plant protection systems. Applicable sensors,

solid state protection system (SSPS) racks, nuclear instrumentation,

and relays will still have response time verified by test prior to

placing the equipment in operational service and after any

maintenance that could affect the response time of that equipment.

Changing the method of periodically verifying instrument response

time for certain instruments from RTT [Response Time Test] to

calibration and channel checks or functional test will not create

any new accident initiators or scenarios. Therefore, the proposed

amendment does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

This change does not affect the total system response time

assumed in the safety analysis. The periodic system response time

verification method for selected pressure and pressure differential

sensors and SSPS racks, nuclear instrumentation, or logic systems is

modified to allow use of actual test data or engineering data

(various Westinghouse WCAPs [topical reports]). The method of

verification still provides assurance that the total system response

time is within that assumed in the safety analysis, since

calibration checks and functional tests will detect any degradation

which might significantly affect equipment response time. Therefore,

the proposed license amendment request does not result in a

significant reduction in margin of safety.

[[Page 54382]]

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Section Chief: Sheri R. Peterson.

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

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

Date of application for amendments: August 30, 1999 (TS 99-10).

Brief description of amendments: The proposed amendments would

change the Sequoyah (SQN) Technical Specifications (TS) to provide

clarification to the requirements for containment isolation valves.

Basis for proposed no significant hazards consideration

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

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

of no significant hazards consideration, which is presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed revisions enhance the technical specification (TS)

requirements to provide greater consistency with the standard TS in

NUREG-1431. This revision proposes changes to the requirements for

containment isolation valves in Specifications 3.6.3. A proposed

revision relocates a surveillance requirement (SR) from SQN TS

3.6.1.1, ``Containment Integrity'' to SQN TS 3.6.3, ``Containment

Isolation Valves.'' A proposed revision to TS 3.6.3, Action (a), a

new Action (b), and a proposed revision to SR 4.6.3.2 provide

improvements to the existing TS requirements. The proposed revisions

are not the result of changes to plant equipment, system design,

testing methods, or operating practices. The modified requirements

will allow some relaxation of current action requirements, and SRs.

These changes provide more appropriate requirements in consideration

of the safety significance and the design capabilities of the plant

as determined by the improved standard TS industry effort. SQN TS

3.6.3, ``Containment Isolation Valves,'' continues to provide

controls to ensure these valves isolate within the time limits

assumed in the safety analyses. Operability of these valves

continues to assure that the containment isolation function assumed

in the safety analyses is maintained. Since these proposed revisions

will continue to support the required safety functions without

modification of the plant features, the probability of an accident

is not increased.

The provisions proposed in this change request will continue to

maintain an acceptable level of protection for the health and safety

of the public and will not significantly impact the potential for

the offsite release of radioactive products. The overall effect of

the proposed change will result in specifications that have

equivalent or improved requirements compared to existing

specifications for containment isolation valve operability and will

not significantly increase the consequences of an accident.

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

new or different kind of accident from any accident previously

evaluated.

The proposed revisions are not the result of changes to plant

equipment, system design, testing methods, or operating practices.

The modified requirements will allow some relaxation of current

action requirements, and a SR consistent with NUREG-1431. These

changes provide more appropriate requirements in consideration of

the safety significance and the design capabilities of SQN's

containment isolation system. The specifications for containment

isolation valves serve to provide controls for maintaining the

containment pressure boundary. TVA's proposed changes does not

contribute to the generation of postulated accidents. Since the

function of the containment isolation valves and their associated

systems remains unchanged, and the effects do not contribute to

accident generation, the proposed changes will not create the

possibility of a new or different kind of accident.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed changes will not result in changes to system design

or setpoints that are intended to ensure timely identification of

plant conditions that could be precursors to accidents or potential

degradation of accident mitigation systems. Operability requirements

for SQN's containment isolation valves remain unchanged. TVA's

proposed revisions provide some relaxation and flexibility to

existing actions and a SR; however, the addition of a new action

requirement for a 31-day periodic verification of valve position

provides conservative administrative controls to ensure containment

isolation function is maintained. The action times are acceptable

considering the redundant features of containment penetration flow

paths and the allowed time intervals that have been developed by the

industry and NRC.

TVA's revisions will continue to provide the necessary actions

to minimize the impact of inoperable containment isolation valves

and will provide testing activities that will ensure containment

isolation system operability. The setpoints and design features that

support the margin of safety are unchanged and actions for

inoperable systems continue to provide appropriate time limits and

compensatory measures. Accordingly, the proposed changes will not

significantly reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Section Chief: Sheri R. Peterson.

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

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

Date of application for amendments: August 30, 1999 (TS 99-11).

Brief description of amendments: The proposed amendments would add

Sequoyah (SQN) Technical Specification (TS) 3.0.7 to address the use of

interim provisions upon discovery of unintended TS action.

Basis for proposed no significant hazards consideration

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

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

of no significant hazards consideration, which is presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

TVA proposes the addition of a new definition and limiting

condition for operation (LCO) that will allow the interim correction

of erroneous TS requirements until NRC's review of an amendment

request is completed. This allowance will only apply to those errors

that are clearly in conflict with the intended purpose of the TS

requirement. The proposed revision will not alter any plant

equipment or operating practices or deviate from the intended

application of the TS requirements. Therefore, the probability of an

accident is not increased by this revision. Likewise, the

consequences of an accident is not increased because the proposed

allowance will maintain the underlying intent of the TS

requirements, the plant licensing basis, and plant nuclear safety.

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

new or different kind of accident from any accident previously

evaluated.

The proposed revision to the SQN TSs will not alter plant

equipment or operating practices. The intent of the TS requirements

will be maintained to ensure the assumed initial conditions for

accidents and the availability of mitigation systems in the event of

an postulated accident. The proposed addition will not promote

activities that have

[[Page 54383]]

the potential to generate accidents. Therefore, the proposed

revision will not create the possibility of an accident of a new or

different kind.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

TVA's proposed revision to add an allowance to correct erroneous

TS requirements will not alter plant systems or those setpoints and

limits that are use[d] to maintain safety functions. Any corrections

implemented in accordance with the proposed allowance will be

consistent with the underlying intent of the TSs. TVA will pursue

timely correction of such errors through the license amendment

process while temporarily utilizing the corrected requirement. This

will ensure that inadequate TS requirements are resolved with NRC in

an acceptable time interval. Implementation of the proposed revision

will enhance the ability to maintain the licensing basis and safety

features of the plant without the need for unnecessary unit

shutdowns or regulatory activities. Therefore, the proposed revision

maintains the plant safety features without the reduction of any

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Section Chief: Sheri R. Peterson

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

Callaway County, Missouri

Date of application request: September 8, 1999.

Description of amendment request: The amendment will authorize

revisions to the Final Safety Analysis Report (FSAR) to reflect

increases in the radiological dose consequences in the Callaway FSAR

for the steam generator tube rupture (SGTR) and main steam line break

(MSLB) accidents.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This change increases the offsite dose consequences for the MSLB

and SGTR accidents reported in FSAR Sections 15.1 and 15.6. Non-

conservative assumptions regarding letdown flow rate, iodine

isotopic mix in the source term, resin effeciency, and termination

of the flash release pathway were identified in the SGTR and MSLB

radiological consequence analyses. The correction of these non-

conservative assumptions results in an increase in the radiological

consequences reported in FSAR Tables 15.1-4 and 15.6-5. However,

these increases are not significant since the new values remain less

than the 10 CFR 100.11 regulatory requirements and the guideline

values provided by the Standard Review Plan [NUREG-0800].

There will be no increase in the probability of previously

evaluated accidents. This change only involves the modeling and

calculation of the SGTR and MSLB radiological consequences. [There

are no equipment or system changes.] Protection system performance

will remain within the assumptions of the previously performed

accident analyses since no hardware changes are proposed. The

protection systems will continue to function in a manner consistent

with the plant design basis. The proposed change will not affect the

probability of any event initiators nor will the proposed change

affect the ability of any safety-related equipment to perform its

intended function. There will be no degradation in the performance

of, nor an increase in the number of challenges imposed on, safety-

related equipment assumed to function during an accident situation.

There will be no change to normal plant operating parameters or

accident mitigation performance.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This change is the result of a re-analysis of the MSLB and SGTR

radiological consequences. These accidents were previously analyzed

in the FSAR. None of the changes in the dose calculation modeling

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

There are no hardware changes associated with this amendment

application nor are there any changes in the method by which any

safety-related plant system performs its safety function. The change

will not affect the normal method of plant operation, other than the

imposition of administrative limits on the concentrations of I-134

[Iodine-134] and Dose Equivalent I-131 until this amendment

application is approved by NRC. No new accident scenarios, transient

precursors, failure mechanisms, or limiting single failures are

introduced as a result of this change. There will be no adverse

effect or challenges imposed on any safety-related system as a

result of this change.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The re-analysis of the MSLB and SGTR radiological consequences,

and the resultant increase in consequences reported in FSAR Tables

15.1-4 and 15.6-5, ensures that the accident analyses support the

plant operating conditions allowed by current Technical

Specification 3.4.8, Reactor Coolant System Specific Activity (ITS

[Improved Technical Specification] 3.4.16), and current Technical

Specification 3.7.1.4, Plant Systems Specific Activity (ITS 3.7.18).

The proposed change does not affect the acceptance criteria for

any analyzed event nor is there a change to any Safety Analysis

Limit (SAL). There will be no effect on the manner in which safety

limits or limiting safety system settings are determined nor will

there be any effect on those plant systems necessary to assure the

accomplishment of protection functions. There will be no impact on

the overpower limit, DNBR [departure from nucleate boiling ratio],

FQ [heat flux hot channel factor], FdeltaH [nuclear

enthalpy rise hot channel factor], LOCA PCT [peak cladding

temperature for the loss-of-coolant accident], peak local power

density, or any other margin of safety. The radiological dose

consequence acceptance criteria listed in the Standard Review Plan

continue to be met.

Therefore, the proposed change does not involve a significant

reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Elmer Ellis Library,

University of Missouri, Columbia Missouri 65201.

Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &

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

NRC Section Chief: Stephen Dembek.

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 54384]]

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: September 14, 1998.

Description of amendment requests: The proposed amendments would

change the runout limits for a safety injection (SI) pump to 675

gallons per minute (gpm), unless the pump is specifically tested to a

higher flow rate, not exceeding 700 gpm for both Units 1 and 2. This

change was initiated upon reevaluation of correspondence from

Westinghouse sent to the licensee in 1991, which indicated that the

generic runout limits for Pacific 2'' JTCH pumps was 675 gpm unless

each specific pump is tested to a higher flow rate. Individual testing

is necessary due to test variations between pumps which may limit the

applicability of testing of one pump to another pump due to

manufacturing tolerances in the sand cast impellers and material

changes in the pump casing.

Furthermore, the bases section is being clarified to describe why

the injection rather than the recirculation mode during flow balancing

is the minimum resistance and, consequently, more conservative

configuration for runout considerations.

Date of publication of individual notice in Federal Register:

August 31, 1999 (64 FR 47533).

Expiration date of individual notice: September 30, 1999

Local Public Document Room location: Maud Preston Palenske Memorial

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

Date of publication of individual notice in Federal Register:

August 31, 1999 (64 FR 47533).

Expiration date of individual notice: September 30, 1999.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendments: October 8, 1998.

Brief description of amendments: The amendments would revise

Technical Specification (TS) 3.3.3.8 for Unit 1 and TS 3.3.3.6 for Unit

2, ``Post-Accident Instrumentation.'' The proposed changes to the TSs

will place tighter restrictions on the amount of time the refueling

water storage tank (RWST) water level instrumentation may be inoperable

before the limiting conditions for operation in the TSs are applied.

Date of publication of individual notice in Federal Register:

August 31, 1999 (64 FR 47532).

Expiration date of individual notice: September 30, 1999.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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 application for amendments: December 3, 1998.

Brief description of amendments: The amendments would make

administrative changes to several Technical Specifications to remove

obsolete information, provide consistency between Unit 1 and Unit 2,

provide consistency with the Standard Technical Specifications, provide

clarification, and correct typographical errors.

Date of publication of individual notice in Federal Register:

August 31, 1999 (64 FR 47535).

Expiration date of individual notice: September 30, 1999.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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 application for amendments: May 21, 1999.

Brief description of amendments: The amendments would change the

Technical Specifications (T/S) to allow reactor coolant system

temperature changes in certain Mode 5 and 6 action statements if the

shutdown margin is sufficient to accommodate the expected temperature

change. In addition, footnotes regarding additions of water from the

refueling water storage tank to the reactor coolant system are

clarified and relocated to action statements. Additional actions are

added in Table 3.3-1, ``Reactor Trip System Instrumentation,'' when the

required source range neutron flux channel is inoperable. Corresponding

changes are proposed for the bases for T/S 3/4.1.1, ``Boration

Control,'' and T/S 3/4.1.2, ``Boration Systems.'' Administrative

changes are proposed to improve clarity. Finally, additions are made to

shutdown margin T/S surveillance requirements to address use of a boron

penalty (requirement for additional boron) during residual heat removal

system operation in Modes 4 and 5.

Date of publication of individual notice in Federal Register: July

12, 1999 (64 FR 37574).

Expiration date of individual notice: August 11, 1999.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

[[Page 54385]]

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

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

Carolina

Date of application for amendment: September 23, 1998, as

supplemented on December 7, 1998, and August 10, 1999.

Brief description of amendment: This amendment revises Technical

Specification (TS) 3/4.6.1.3, ``Containment Air Locks,'' and its

associated bases, to clarify the requirements for locking an air lock

door shut and to make it consistent with NUREG-1431, Revision 1,

``Standard Technical Specifications, Westinghouse Plants,'' dated April

1995.

Date of issuance: September 14, 1999.

Effective date: September 14, 1999.

Amendment No.: 90.

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

Technical Specifications

Date of initial notice in Federal Register: October 21, 1998 (63 FR

56239)

The December 7, 1998, and August 10, 1999, submittals contained

clarifying information only, and did not change the initial no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 14, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

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

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

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

Carolina

Date of application for amendment: June 15, 1999.

Brief description of amendment: This amendment changes the

Technical Specifications to incorporate the performance-based 10 CFR 50

Appendix J, Option B for Type A tests (containment integrated leakage

rate tests). Option B will be implemented for Type A testing in

accordance with NRC Regulatory Guide 1.163, ``Performance-Based

Containment Leak-Test Program,'' dated September 1995, and Nuclear

Energy Institute (NEI) Guideline 94-01, Revision 0, ``Industry

Guideline for Implementing Performance-Based Option of 10 CFR Part 50,

Appendix J,'' dated July 26, 1995. Type B and C testing (containment

penetration leakage tests) will continue to be performed in accordance

with 10 CFR 50 Appendix J, Option A.

Date of issuance: September 17, 1999.

Effective date: September 17, 1999.

Amendment No.: 91.

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

Technical Specifications.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38023). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

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

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of application for amendments: July 30, 1999.

Brief description of amendments: The amendments changed the maximum

allowable temperature of the ultimate heat sink in the technical

specifications from 98 degrees Fahrenheit to 100 degrees Fahrenheit.

The change is in effect from the date of this amendment until September

30, 1999.

Date of issuance: September 8, 1999.

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

Amendment Nos.: 103 and 103.

Facility Operating License Nos. NPF-72 and NPF-77: The amendments

revised the Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: Yes (64 FR 44962 dated August 18, 1999). The notice

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

no significant hazards consideration determination. No comments have

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

hearing by September 17, 1999, but indicated that if the Commission

makes a final no significant hazards consideration determination any

such hearing would take place after issuance of the amendments. The

Commission's related evaluation of the amendments, finding of exigent

circumstances and final no significant hazards consideration

determination are contained in a Safety Evaluation dated September 8,

1999.

Local Public Document Room location: Wilmington Public Library, 201

S. Kankakee Street, Wilmington, Illinois 60481.

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

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

Date of application for amendments: May 3, 1999, as supplemented by

letter dated September 10, 1999.

Brief description of amendments: The amendments relocated the

requirements of Technical Specification (TS) Section 3/4.6.I to the

Updated Final Safety Analysis Report (UFSAR). TS Section

3/4.6.I contains reactor coolant chemistry limiting conditions for

operation (LCO) and surveillance requirements (SR) for conductivity,

chloride concentration, and pH.

Date of issuance: September 23, 1999.

Effective date: Immediately, to be implemented within 30 days

including relocation of the removed TSs and associated bases to the

licensee's UFSAR pending change file. In addition, the licensee shall

include the relocated information in the UFSAR submitted to the NRC,

pursuant to 10 CFR 50.71(e), except for any information that has been

changed in accordance with 10 CFR 50.59 and described in the change

summaries submitted to NRC pursuant to 10 CFR 50.59.

Amendment Nos.: 173 & 169.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: August 11, 1999 (64 FR

43768). The September 10, 1999, submittal provided additional

clarifying information that did not change the initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 23, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

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

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

Date of application for amendments: June 29, 1999.

Brief description of amendments: The amendments increased the notch

testing surveillance interval of partially withdrawn control rods in

Technical Specification Surveillance Requirement 3/4.3.C, ``Reactivity

Control--Control Rod Operability,'' from an interval of once in 7 days

to once in 31 days.

Date of issuance: September 23, 1999.

[[Page 54386]]

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

Amendment Nos.: 190 & 187.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 28, 1999 (64 FR

40905).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 23, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

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

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

Date of application of amendments: May 24, 1999

Brief description of amendments: The amendments revise the maximum

local fuel pin centerline temperature safety limit in Technical

Specification 2.1.1.1 from the limit determined using the TACO2 fuel

performance computer code to the value determined using a newer TACO3

computer code.

Date of Issuance: September 24, 1999.

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

within 30 days from the date of issuance.

Amendment Nos.: Unit 1--306, Unit 2--306, Unit 3--306.

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

Amendments revised the Technical Specifications.

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

35203).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 24, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

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 application for amendments: July 22, 1998, and supplemented

by letters dated October 22, 1998, January 28, May 6, June 24, August

17 and September 15, 1999.

Brief description of amendments: The amendments revise various

sections of the Technical Specifications (Appendix A of the Catawba

operating licenses) to permit use of Westinghouse's Robust Fuel

Assemblies for future core reloads.

Date of issuance: September 22, 1999.

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

prior to beginning the installation of the Westinghouse fuel, currently

projected to be Fuel Cycle 13 and 11 for Units 1 and 2, respectively.

Amendment Nos.: Unit 1--180; Unit 2--172.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64108); May 19, 1999 (64 FR 27317); August 11, 1999 (64 FR 43770)

The Commission's related evaluation of the amendments is contained in a

Safety Evaluation dated September 22, 1999.

No significant hazards consideration comments received: No

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

Black Street, Rock Hill, South Carolina.

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

Nuclear Station, Units 1 and 2, Mecklenberg County, North Carolina

Date of application for amendments: July 22, 1998, and supplemented

by letters dated October 22, 1998, and January 28, May 6, June 24,

August 17 and September 15, 1999

Brief description of amendments: The amendments revise various

sections of the Technical Specifications (Appendix A of the McGuire

operating licenses) to permit use of Westinghouse's Robust Fuel

Assemblies for future core reloads.

Date of issuance: September 22, 1999.

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

prior to beginning the installation of the Westinghouse fuel, currently

projected to be Fuel Cycle 15 and 14 for Units 1 and 2, respectively.

Amendment Nos.: Unit 1--188; Unit 2--169.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: August 11, 1999 (64 FR

43771); June 30, 1999 (64 FR 35202); December 16, 1998 (64 FR 69388)

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 22, 1999.

No significant hazards consideration comments received: No

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

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina

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

No. 1, Pope County, Arkansas

Date of amendment request: April 9, 1999, as supplemented by letter

dated July 29, 1999

Brief description of amendment: The amendment revises the

requirements associated with the station batteries and the direct

current (DC) sources to the 125 volt DC switchyard distribution system.

Date of issuance: September 14, 1999.

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

within 45 days from the date of issuance (including issuance of the

Technical Requirements Manual for use by licensee personnel).

Amendment No.: 200.

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

Technical Specifications.

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

27321).

The July 29, 1999, letter provided clarifying and additional

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 14, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

No. 1, Pope County, Arkansas

Date of amendment request: June 1, 1999, as supplemented by letters

dated July 29 and August 19, 1999.

Brief description of amendment: The amendment revised the Technical

Specifications to allow, under specific conditions, certain once-

through steam generator (OTSG) tubes with tube end crack indications

adjacent to the primary cladding region of the upper and lower OTSG

tubesheets to remain in service.

Date of issuance: September 14, 1999.

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

prior to reactor startup after refueling outage 1R15.

Amendment No.: 201.

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

Technical Specifications.

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

35205).

The July 29 and August 19, 1999, letters provided clarifying

information

[[Page 54387]]

that did not change the scope of the June 1, 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 September 14, 1999.

No significant hazards consideration comments received: No

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit 3, Citrus County, Florida

Date of application for amendment: May 17, 1999.

Brief description of amendment: The amendment changes Technical

Specification Section 3.3.8, ``Emergency Diesel Generator Loss of Power

Start,'' Surveillance Requirement 3.3.8.1 and corresponding basis

section. The surveillance is revised to make a note included in the

surveillance consistent with the method of performing the surveillance.

Date of issuance: September 13, 1999.

Effective date: September 13, 1999.

Amendment No.: 187.

Facility Operating License No. DPR-72: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38026).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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

Generating Station, Ocean County, New Jersey

Date of application for amendment: December 23, 1998.

Brief description of amendment: The proposed amendment revised the

surveillance frequency for verifying the operability of motor-operated

isolation valves and condensate makeup valves in the Isolation

Condenser Technical Specification 4.8.A.1 and Bases page from once per

month to once per 3 months.

Date of Issuance: September 24, 1999.

Effective date: Date of issuance and shall be implemented within 30

days of issuance.

Amendment No.: 209.

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

Technical Specifications.

Date of initial notice in Federal Register: April 7, 1999 (64 FR

17026).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated September 24, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

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

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: October 19, 1998, as

supplemented August 19, 1999.

Brief description of amendment: The proposed amendment adds

operability and surveillance requirements to the Technical

Specifications for the remote shutdown system similar to the standard

technical specifications for Babcock & Wilcox nuclear plants as

described in NUREG-1430.

Date of issuance: September 22, 1999.

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

30 days.

Amendment No.: 216.

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64118). The August 19, 1999, supplement to the application did not

change the staff's proposed no significant hazards consideration

determination or expand the scope of the application as originally

noticed.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 22, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (Regional Depository) Walnut

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

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

County, Iowa

Date of application for amendment: April 30, 1999.

Brief description of amendment: The amendment revises Duane Arnold

Energy Center (DAEC) Technical Specification (TS) Surveillance

Requirement (SR) 3.4.3.1 to revise the safety function lift setpoint

tolerance limits for the main safety valves (SVs) and the safety/relief

valves (SRVs).

Date of issuance: September 22, 1999.

Effective date: September 22, 1999, to be implemented within 30

days.

Amendment No.: 228.

Facility Operating License No. DPR-49: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38028).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 22, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401.

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: May 15, 1998, as supplemented by

letters dated September 25, October 13, December 9 (two letters), 1998;

January 11, April 1, and April 22, 1999.

Brief description of amendment: This amendment changes Technical

Specification (TS) 5.5, ``Storage of Unirradiated and Spent Fuel,'' to

reflect a planned modification to increase the storage capacity of the

spent fuel pool from 2776 to 4086 fuel assemblies. It also deletes an

inappropriate statement and reference within TS 5.5.

Date of issuance: June 17, 1999.

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

of its issuance to be implemented before spent fuel is stored within

the new high-density spent fuel rack modules authorized for

installation and use by this amendment.

Amendment No.: 167.

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

Technical Specifications.

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

FR 64973).

The September 25, October 13, December 9 (two letters) 1998,

January 11, April 1, and April 22, 1999, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

[[Page 54388]]

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

York 13126.

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

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

Date of amendment request: June 23, 1999.

Description of amendment request: To revise Technical Specification

(TS) 3.7.6.2 to increase the allowable outage time for the Control Room

Air Conditioning Subsystem from 30 days to 60 days, on a one-time basis

for each train, to allow adequate time to replace portions of the

existing system during the current operating cycle, and to exclude the

requirements of TS 3.0.4 and TS 4.0.4 during the implementation of the

modification.

Date of issuance: September 17, 1999.

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

implemented within 30 days.

Amendment No.: 62.

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

Technical Specifications/License.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38032).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

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

Date of application for amendment: March 17, 1999.

Brief description of amendment: The amendment changes Technical

Specifications 3.5.2, ``Emergency Core Cooling Systems--ECCS

Subsystems--Tavg 300 deg.F;'' 3.7.1.7, ``Plant Systems--

Atmospheric Steam Dump Valves;'' and 3.7.6.1, ``Plant Systems--Control

Room Emergency Ventilation System.'' The changes will revise: (1)

Surveillance requirements for the Emergency Core Cooling System valves,

(2) the atmospheric steam dump valve requirements to focus on the steam

release path instead of the individual valves, and (3) the allowed

outage time for the atmospheric steam valves and Control Room Emergency

Ventilation System. The licensee made changes to the Bases pages

consistent with the proposed changes to the TSs.

Date of issuance: August 12, 1999.

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

within 60 days from the date of issuance.

Amendment No.: 238.

Facility Operating License No. DPR-65: Amendment revised the

Technical Specifications.

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

19559).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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 application for amendment: June 4, 1999.

Brief description of amendment: The amendment makes administrative

changes to the Technical Specifications.

Date of issuance: September 14, 1999.

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

30 days.

Amendment No.: 193.

Facility Operating License No. DPR-64: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: July 28, 1999 (64 FR

40906).

No significant hazards consideration comments received: No.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 14, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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 application for amendment: January 28, 1999, as

supplemented April 29, 1999, and May 17, 1999. By letters dated April

29, 1999, and May 17, 1999, the licensee revised the original submittal

dated January 28, 1999, in response to questions raised by the NRC

staff.

Brief description of amendment: The amendment changes the Technical

Specifications by reducing the number of emergency diesel generators

required to be operable under certain conditions.

Date of issuance: September 14, 1999.

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

30 days.

Amendment No.: 194.

Facility Operating License No. DPR-64: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 2, 1999 (64 FR

29713). This notice superceded a notice dated April 21, 1999 (64 FR

19563).

No significant hazards consideration comments received: No.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 14, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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 application for amendment: January 28, 1999, as

supplemented July 16, 1999.

Brief description of amendment: The amendment removes lists of

containment isolation valves from the Technical Specifications (TSs)

and modifies the TSs accordingly.

Date of issuance: September 16, 1999.

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

30 days.

Amendment No.: 195.

Facility Operating License No. DPR-64: Amendment revises the

Technical Specifications.

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

27323).

The July 16, 1999, submittal did not change the staff's initial

proposed finding of no significant hazards considerations.

No significant hazards consideration comments received: No.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

[[Page 54389]]

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: April 5, 1999.

Brief description of amendment: The proposed changes would revise

Appendix A (Section 6.1) and Appendix B (Section 7.1) of the James A.

FitzPatrick Technical Specifications. The proposed changes would remove

the position title of General Manager from these sections and would

state that if the Site Executive Officer is unavailable, he will

delegate his responsibilities to another staff member, in writing. In

addition the position title of Resident Manager, used in Appendix B,

Section 7.1, would be replaced by Site Executive Officer.

Date of issuance: September 13, 1999.

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

30 days.

Amendment No.: 254.

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

Technical Specifications and the Environmental Technical

Specifications.

Date of initial notice in Federal Register: August 11, 1999 (64 FR

43775).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: October 8, 1997.

Brief description of amendment: The amendment revises actions in

the Technical Specifications to be taken in the event multiple control

rods are inoperable.

Date of issuance: September 21, 1999.

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

30 days.

Amendment No.: 255.

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

Technical Specifications.

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

FR 6991).

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.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

Generating Station, Salem County, New Jersey

Date of application for amendment: December 30, 1998, as

supplemented September 13, 1999.

Brief description of amendment: This amendment revises Technical

Specification (TS) Limiting Condition for Operation 3.7.3 and TS Table

3.7.3-1. These changes modify the flood protection actions required

when severe storm warnings that may affect the site are in effect or

during periods of elevated river water level.

Date of issuance: September 17, 1999.

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

implemented within 60 days.

Amendment No.: 122.

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

TSs.

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

FR 9200).

The September 13, 1999, supplement provided clarifying information

that did not change the initial proposed no significant hazards

determination or expand the scope of the initial Federal Register

notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

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

Generating Station, Salem County, New Jersey

Date of application for amendment: May 24, 1999, as supplemented

June 21, 1999.

Brief description of amendment: This amendment revises the

Technical Specifications (TSs) to correct typographical and editorial

errors, and is considered administrative in nature.

Date of issuance: September 21, 1999

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

implemented within 60 days.

Amendment No.: 123.

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

TSs.

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

35209).

The June 21, 1999, supplement 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 amendment is contained

in a Safety Evaluation dated September 21, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

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: July 2, 1999.

Brief description of amendments: The amendments delete TS 3/4.3.4,

``Instrumentation--Turbine Overspeed Protection,'' and its associated

Bases and relocate the requirements to the licensee-controlled Updated

Final Safety Analysis Report.

Date of issuance: September 14, 1999

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

within 60 days.

Amendment Nos.: 224 and 205.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: August 11, 1999 (64 FR

43776).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 14, 1999.

No significant hazards consideration comments received: No.

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

West Broadway, Salem, NJ 08079.

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: December 31, 1998 (PCN-501), as

supplemented June 14, 1999.

Brief description of amendments: The amendments consist of changes

to Technical Specification 3.3.5, ``Engineered Safety Features

Actuation System (ESFAS) Instrumentation,'' and will include

restrictions on operation with a channel of the refueling water storage

tank level--low input to the recirculation actuation signal and the

steam generator pressure--low input or

[[Page 54390]]

steam generator pressure difference--high input to the emergency

feedwater actuation signal in the tripped condition.

Date of issuance: September 7, 1999.

Effective date: September 7, 1999, to be implemented within 30 days

of issuance.

Amendment Nos.: Unit 2--157; Unit 3--148.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 28, 1999 (64 FR

40907).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 7, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

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

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: June 18, 1997 (PCN-478), as

supplemented May 24 and August 10, 1999.

Brief description of amendments: The amendments modify the

Technical Specification surveillance requirements related to diesel

generator testing to more clearly reflect safety analysis and testing

conditions as it is performed.

Date of issuance: September 9, 1999.

Effective date: September 9, 1999, to be implemented within 30 days

of issuance.

Amendment Nos.: Unit 2--158; Unit 3--149.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: December 31, 1997 (62

FR 68315) The licensee's letters dated May 24 and August 10, 1999,

provided updated Technical Specification pages, clarifications, and

additional information that were 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 amendments is contained

in a Safety Evaluation dated September 9, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Main Library, University of

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

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

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

Date of amendment request: June 7, 1999.

Brief description of amendments: The amendments revised Technical

Specification (TS) 2.2.1, ``Reactor Trip System (RTS) Instrumentation

Setpoints,'' and TS 3.3.2, ``Engineered Safety Features Actuation

System (ESFAS) Instrumentation,'' and the associated Bases, by removing

the Total Allowance, Sensor Error, and Z terms (Z is the statistical

summation of errors excluding sensor and rack drift) from the RTS and

ESFAS Instrumentation Trip Setpoints Tables. This replaces the five-

column methodology with a two-column methodology that consists of the

trip setpoint and allowable value columns.

Date of issuance: September 13, 1999.

Effective date: September 13, 1999, to be implemented within 30

days.

Amendment Nos.: Unit 1--116; Unit 2--104.

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

revised the Technical Specifications.

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

35211) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

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

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

Date of application for amendments: June 24, 1999 (TS 99-06).

Brief description of amendments: The amendments revise the Sequoyah

Nuclear Plant Technical Specifications (TS) by adding a footnote to

allow use of an installed spare electrical inverter, if needed.

Date of issuance: September 23, 1999.

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

later than 45 days after issuance.

Amendment Nos.: 246 and 237.

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

revise the TS.

Date of initial notice in Federal Register: August 2, 1999 (64 FR

41973) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 23, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

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

Date of amendment request: June 23, 1999, as supplemented by letter

dated August 4, 1999.

Brief description of amendments: The amendments revise Surveillance

Requirement 3.8.1.13, ``AC Sources--Operating'' to clarify that each

emergency diesel generator automatic noncritical trip, except for

engine overspeed and generator differential current, is bypassed on

either a loss-of-offsite power or a safety injection actuation signal.

Date of issuance: September 21, 1999.

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

within 30 days from the date of issuance.

Amendment Nos.: 69 and 69.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38037) The August 4, 1999, letter provided additional and clarifying

information that did not change the scope of the June 23, 1999,

application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 21, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas 76019.

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

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

Date of amendment request: February 12, 1999, as supplemented by

letter dated June 14, 1999

Brief description of amendments: The amendments change Technical

Specification (TS) 3.4.13, ``RCS [Reactor Coolant System] Operational

Leakage,'' TS 5.5.9, ``Steam Generator (SG) Tube Surveillance

Program,'' and TS 5.6.10,

[[Page 54391]]

``Steam Generator Tube Inspection Report,'' to implement the 1.0 Volt

Steam Generator Tube Repair Criteria for CPSES, Unit 1.

Date of issuance: September 22, 1999.

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

within 30 days from the date of issuance.

Amendment Nos.: Unit 1--Amendment No. 70; Amendment No. 70.

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

revised the Technical Specifications.

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

24202) The June 14, 1999, supplement provided clarifying information

that did not change the scope of the February 12, 1999, application and

the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 22, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas 76019.

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

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

Date of amendment request: October 2, 1998, as supplemented by

letters dated July 27 and August 26, 1999.

Brief description of amendments: The amendments revise Technical

Specfications for CPSES, Unit 1, to define the F* steam generator tube

plugging criteria in TS 5.5.9, ``Steam Generator (SG) Tube Surveillance

Program,'' and associated reporting requirements in TS 5.6.10, ``Steam

Generator Inspection Report.''

Date of issuance: September 22, 1999.

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

within 30 days from the date of issuance.

Amendment Nos.: Unit 1--Amendment No. 71; Unit 2--Amendment No. 71.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 4, 1998 (63 FR

59597). The July 27 and August 26, 1999, letters provided clarifying

information that did not change the scope of the October 2, 1998,

application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 22, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas 76019.

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

Yankee Nuclear Power Station, V

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