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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from November 20, 1999, through December 3, 1999.

The last biweekly notice was published on December 1, 1999 (64 FR

67330).

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Review and Directives Branch, Division of Freedom of Information and

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

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

publication date and page number of this Federal Register notice.

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

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

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

[[Page 70078]]

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

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

petitions for leave to intervene is discussed below.

By January 14, 2000, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

Building, 2120 L Street, NW., Washington, DC, and electronically from

the ADAMS Public Library component on the NRC Web site, http://

www.nrc.gov (the Electronic Reading Room). If a request for a hearing

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

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

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

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

the designated Atomic Safety and Licensing Board will issue a notice of

a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

operator should be given Datagram Identification Number N1023 and the

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

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

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

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

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

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

Street, NW., Washington, DC, and electronically from the ADAMS Public

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

Electronic Reading Room)

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs

Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

Date of amendment request: November 18, 1999.

Description of amendment request: The proposed amendment revises

the Unit 1 Heatup Curve (Technical Specification Figure 3.4.3-1), Unit

1 Cooldown Curve (Technical Specification Figure 3.4.3-2), and Unit 1

Maximum Power-Operated Relief Valve (PORV) Opening Pressure vs

Temperature Curve (Technical Specification Figure 3.4.12-1) to change

fluence level from 2.61 x 10\19\ n/cm \2\ to 4.49 x 1019 n/cm \2\

(E>1MeV). This change reflects the new actual fluence level for which

these curves are valid, and is necessary to extend the

[[Page 70079]]

applicability of the curves for Unit 1 operation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

In accordance with 10 CFR Part 50, Appendix G, the Calvert

Cliffs pressure/temperature (P-T) limits for material fracture

toughness requirements of the reactor coolant pressure boundary

materials were developed using the methods of linear elastic

fracture mechanics and the guidance found in the American Society of

Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section

III, Appendix G. The Calvert Cliffs (P-T) limits are based on

fluence level. The fluence level corresponds to the pressurized

thermal shock (PTS) screening criteria defined in 10 CFR 50.61 for

the critical elements. Methods described in the Nuclear Regulatory

Commission Regulatory Guide 1.99, Revision 2, are used to predict

the embrittlement effect of neutron irradiation on reactor vessel

materials. Regulatory Guide 1.99 defines embrittlement effect in

terms of adjusted reference temperatures (ART), which depends on the

material property of the PTS critical element.

The proposed higher fluence level for the Technical

Specification P-T limits was made possible by the identification of

a new 10 CFR 50.61 critical element for fracture toughness

requirements for protection against PTS events. The material

properties of the new critical element resulted in an increase in

fluence level from 2.61 x 10 \19\ n/cm \2\ to 4.49 x 1019 n/cm \2\

for the ART valves calculated using the material properties of the

old PTS critical element. the P-T limits analysis remain well within

the conservative acceptance limits of the ASME Boiler and Pressure

Vessel Code, Section III, Appendix G. Hence, with the new higher

fluence level, the 10 CFR Part 50, Appendix G, requirement for

adequate margin to brittle failure during normal operation,

anticipated operational occurrences, and system hydrostatic tests,

for the reactor coolant pressure boundary materials, is maintained.

Therefore the proposed change 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 any accidents previously evaluated.

The implementation of the proposed revision has no significant

effect on either the configuration of the plant, or the manner in

which it is operated.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

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

safety.

As discussed above, the P-T limits analysis remain well within

the conservative acceptance limits of the ASME Boiler and Pressure

Vessel Code, Section III, Appendix G. Hence, with the new higher

fluence level, the 10 CFR Part 50, Appendix G, requirement for

adequate margin to brittle failure during normal operation,

anticipated operational occurrences, and system hydrostatic tests,

for the reactor coolant pressure boundary materials, is maintained.

Therefore, this proposed modification does not significantly

reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

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

NRC Acting Section Chief: Victor Nerses.

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: November 19, 1999.

Description of amendments request: The amendments request approval

of changes in the Updated Final Safety Analysis Report (UFSAR) that

constitute an unreviewed safety question (USQ) as described in 10 CFR

50.59. Specifically, these changes would be an increase in the

probability of occurrence of malfunction. Additionally, these changes

were not previously evaluated in the UFSAR.

Regulations require that structures, systems, and components

important to safety be appropriately protected against the effects of

effects of missiles that might result from equipment failures. Failures

that could occur in the large turbines of the main turbine-generator

sets have the potential for producing large high-energy missiles

(hereinafter called ``turbine missiles''). Both of Baltimore Gas and

Electric Company's (BGE) turbine generator suppliers studied the

failure of the rotating elements of their turbine-generators. The UFSAR

only addresses a turbine missile hitting the Containment Building,

Control Room, Switchgear Room, and Waste Processing Area. As a result

of revising the Unit 1 and Unit 2 turbine missile analysis, BGE

determined that the discussion of turbine missiles in Section 5.3.1 of

the UFSAR was incomplete. Specifically, it did not discuss the

probability of a missile from the Unit 1 turbine-generator striking: 1)

the refueling water tanks; 2) the No. 11 Fuel Oil Storage Tank; or 3)

plant equipment through various roof slabs or through non-missile-proof

openings in the missile-proof walls. When these additional targets are

included, the total target area is increased. If the target area

increases, the probability of a turbine missile causing equipment

damage increases. It is this increase in probability that leads to a

USQ for a turbine missile from Unit 1. Note that by using methodologies

previously approved by NRC, the revised analysis concludes there is no

USQ for turbine missiles from the Unit 2 turbine-generator.

The UFSAR change is considered a USQ for Units 1 and 2 because the

results of the revised Unit 1 turbine missile analysis for the

following unprotected rooms or components show an increase in

probability of occurrence of malfunction not previously evaluated in

the UFSAR:

the Refueling Water Tanks;

the No. 11 Fuel Oil Storage Tank (non-missile-proof);

the saltwater pumps through roof hatches in the Intake Structure

roof;

the roof slabs over the refueling Water Tank Pump Room, the Control

Room Heating, Ventilation, and Air Conditioning (HVAC) Equipment Room,

the Spent Fuel Pool Area Ventilation Equipment room, and a portion of

118 level roof over the fuel cask handling area;

the Control Room HVAC Room through its non-missile-proof door; and

the Unit 1 Auxiliary Building 45 Switchgear Room through

the its non-missile-proof doors.

The probability of a missile from the Unit 1 turbine-generator

striking them is a negligible increase in the probability of occurrence

of malfunction of equipment associated with Unit 1 and 2. Upon approval

of this request, the UFSAR will be revised to reflect the proposed

turbine missile description. There is no USQ associated with the Unit 2

turbine-generator.

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.

Regulations require that structures, systems, and components

important to safety be appropriately protected against the effects

of missiles that might result from equipment failures. Further that

could occur in the large turbines of the main turbine-generator sets

have the potential for producing large high-

[[Page 70080]]

energy missiles (hereinafter called turbine missiles). Both of our

turbine-generator suppliers studied the failure of the rotating

elements of their turbine-generators. The UFSAR only addresses

turbine missile hitting the Containment Building, Control Room,

Switchgear Room, and Waste Processing Area. As result of revising

the Unit 1 and Unit 2 turbine missile analysis, we determined that

the discussion of turbine missiles of the UFSAR was incomplete. From

the revised analysis, we determined Unit 1 and 2 USQs exist for the

following unprotected rooms or components (i.e., there is an

increase in probability of occurrence of malfunction not previously

evaluated in the UFSAR):

the Refueling Water Tanks;

the No. 11 Fuel Oil Storage Tank;

the Saltwater Pumps through roof hatches in the Intake Structure

Roof;

the roof slabs over the Refueling Water Tank Pump Room, the

Control Room Heating, Ventilation, and Air Conditioning (HVAC)

Equipment Room, Spent Fuel Pool Area Ventilation Equipment Room, and

a portion of 118' level roof over the cask handling area;

the Control Room HVAC Room through its non-missile-proof door;

and,

the Unit 1 Auxiliary Building 45' Switchgear Room through its

non-missile-proof doors.

The probability of a missile from the Unit 1 turbine-generator

striking them is a negligible, but greater than zero, increase in

the probability of occurrence of malfunction of equipment associated

with Units 1 and 2.

For Unit 1 High Trajectory Missiles (HTM), the guidance of NUREG

0800, Standard Review Plan, is used as one acceptable method for

evaluating the risk. Use of this method is not a commitment to the

Standard Review Plan and does not incorporate the Standard Review

Plan into our licensing basis. The revised analysis shows that the

total target area considered vulnerable to an HTM is less than the

Standard Review Plan limit of 10,000 ft2 for each unit.

Therefore, the risk form an HTM is insignificant. Note that all of

the Units 1, 2, and Common structures listed above are equally

vulnerable to a Unit 1 HTM. Therefore, any risk increase to the

plant structures constitutes a USQ for Units 1 and 2.

For Unit 1 Low-Trajectory Missiles (LTMs), protection for the

Auxiliary Building is provided by a 3' thick, concrete, missile-

proof wall between the Turbine Building and the Auxiliary building

(the K-line wall). This wall is 3' thick below the 69' elevation and

2' thick above the 69' for areas protecting safety-related

equipment. The revised analysis evaluates the protection of Unit 1

equipment from a Unit 1 LTM. The 69' Control Room HVAC Equipment

Room and Unit 1 Auxiliary Building 45' Switchgear Room are protected

by the missile-proof walls except for the openings at the non-

missile-proof doors. A turbine missile that hits one of these doors

is assumed to go through them, strike safety-related equipment in

the room, and cause it to fail. Recall that the Control Room HVAC

equipment is shared by both units. Therefore, any increase in risk

of failure of equipment in this room affects both Units 1 and 2.

The risk associated with a turbine missile to either of these

doors is calculated using guidance in Regulatory Guide 1.115,

Revision 1, ``Protection Against Low-Trajectory Turbine Missiles.''

This guidance states that the turbine missile hazard should be less

than 107. The missile hazard rate in the revised risk

analysis shows that the risk from LTMs from the Unit 1 General

Electric turbine-generator to the 69' Control Room HVAC Equipment

Room and Unit 1 Auxiliary Building 45' Switchgear room through these

non-missile-proof doors is less than 107.

Based on the above, the proposed change 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 any accident previously evaluated.

The proposed change makes no physical changes to the plant.

Specifically, the proposed change does not add new or modify

existing plant equipment such that it could become an accident

initiator different from its current role as an accident initiator.

The only change made by this activity is the revision of the UFSAR

to include the revised turbine missile analysis. The UFSAR chapter 1

drawings correctly depict the location of plant structures and

components, including the thickness of and the openings in the

missile-proof wall between the Turbine Building and the Auxiliary

building (the K-Line Wall). Therefore, the possibility of a new or

different type of accident is not created by the proposed change.

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

safety.

The regulations require an evaluation of turbine missiles to

ensure that structures, systems, and components important to safety

be appropriately protected from them. Revised turbine missile

analysis have been performed consistent with appropriate regulatory

guidance (Regulatory Guide 1.115 and the Standard Review Plan). The

results of the revised analysis meet the acceptance criteria of the

guidance. 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

amendments request involves no significant hazards consideration.

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

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

NRC Acting Section Chief: Victor Nerses.

Carolina Power & Light Company, et al., Docket No. 50-325, Brunswick

Steam Electric Plant, Unit 1, Brunswick County, North Carolina

Date of amendment request: November 17, 1999.

Description of amendment request: The proposed amendment would

change Technical Specification (TS) 2.1.1.2, ``Reactor Core Safety

Limits.'' The minimum critical power ratios (MCPR) for single and two

recirculation loop operation would be increased. In addition, the

reference in TS 5.6.5, ``Core Operating Limits Report,'' Item b.5,

would be removed.

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 license amendments do not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

The proposed license amendment will establish MCPR Safety Limit

values of 1.10 for two recirculation loop operation and 1.11 for single

recirculation loop operation. Additionally, the proposed license

amendment replaces an expiring cycle-specific reference in the list of

analytical methods approved for determining core operating limits in

Specification 5.6.5.b with a reference to a GE [General Electric]

topical report which has been accepted by the NRC.

The methods for calculating the MCPR Safety Limit values have been

previously approved by the NRC and are described in GE's reload

licensing methodology topical report NEDE-24011-P-A. Use of these

methods ensures that the integrity of the fuel will be maintained

during normal operation and that the resulting MCPR Safety Limit values

satisfy the fuel design safety criteria that less than 0.1 percent of

the fuel rods experience boiling transition if the safety limits are

not violated. The change does not require any physical plant

modifications, physically affect any plant components, or allow the

plant to be operated any closer to fuel design limits. Therefore, the

proposed change to the MCPR Safety Limit values and to the list in

Specification 5.6.5.b of analytical methods approved for determining

core operating limits results no increase in the probability of a

previously evaluated accident.

The consequences of a previously evaluated accident are dependent

on the initial conditions assumed for the analysis, the behavior of the

fuel during the accident, the availability and successful functioning

of the equipment assumed to operate in response to the accident, and

the setpoints at which these actions are initiated.

[[Page 70081]]

The methods used for calculating the MCPR Safety Limits have been

approved by the NRC and are described in GE's reload licensing

methodology topical report NEDE-24011, ``General Electric Standard

Application for Reactor Fuel (GESTAR II).'' The proposed MCPR Safety

Limit values of 1.10 for two recirculation loop operation and 1.11 for

single recirculation loop operation will ensure that less than 0.1

percent of the fuel rods will experience boiling transition during any

plant operation if the limits are not violated. The proposed change to

the MCPR Safety Limit values does not affect the performance of any

equipment used to mitigate the consequences of a previously evaluated

accident. Also, the proposed change does not affect setpoints that

initiate protective or mitigative actions. No analysis assumptions are

violated and there are no adverse effects on the factors contributing

to offsite and onsite dose.

Based on the determination of the proposed MCPR Safety Limit values

using conservative NRC-approved methods and the operability of plant

systems designed to mitigate the consequences of accidents not being

changed, the proposed change to the MCPR Safety Limit values and to the

list in Specification 5.6.5.b of analytical methods approved for

determining core operating limits does not significantly increase the

consequences of a previously evaluated accident.

2. The proposed license amendments will not create the

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

previously evaluated.

Creation of the possibility of a new or different kind of

accident would require the creation of one or more new precursors of

that accident. New accident precursors may be created by

modifications of the plant configuration, including changes in

allowable modes of operation. This proposed license amendment does

not involve any physical alteration of plant systems and plant

equipment will not be operated in a different manner. As a result,

no new failure modes are being introduced. Therefore, the proposed

change to the MCPR Safety Limit values and to the list in

Specification 5.6.5.b of analytical methods approved for determining

core operating limits will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed license amendments do not involve a significant

reduction in a margin of safety.

The margin of safety is established through the design of the

plant structures, systems, and components; through the parameters

within which the plant is operated; through the establishment of

setpoints for actuation of equipment relied upon to respond to an

event; and through margins contained within the safety analyses.

The proposed change to the MCPR Safety Limit values and the list

in Specification 5.6.5.b of analytical methods approved for

determining core operating limits does not adversely impact the

performance of plant structures, systems, components, and setpoints

relied upon to respond to mitigate an accident. As previously

stated, the methods for calculating the MCPR Safety Limit values

have been previously approved by the NRC and are described in GE's

reload licensing methodology topical report NEDE-24011-P-A. Use of

these methods ensures that the resulting MCPR Safety Limit values

satisfy the fuel design safety criteria that less than 0.1 percent

of the fuel rods experience boiling transition if the safety limits

are not violated. As a result, the proposed changes do not

significantly impact any safety analysis assumptions or results.

Based on the assurance that the fuel design safety criteria will be

met, the proposed changes do not involve a significant reduction in

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602.

NRC Section Chief: Richard P. Correia.

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: November 19, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) for the Harris Nuclear Plant

(HNP) to incorporate American Society for Testing and Materials (ASTM)

D3803-1989, ``Standard Test Method for Nuclear-Grade Activated

Carbon,'' as the standard for testing nuclear-grade activated charcoal.

Specifically, TS 4.7.6 will be revised for the Control Room Emergency

Filtration System, TS 4.7.7 will be revised for the Reactor Auxiliary

Building Emergency Exhaust System, and TS 4.9.12 will be revised for

the Fuel Handling Building Emergency Exhaust System. These changes are

being proposed in accordance with NRC Generic Letter (GL) 99-02,

``Laboratory Testing Of Nuclear-Grade Activated Charcoal,'' dated June

3, 1999.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

This proposed change to revise the standard to which activated

charcoal samples are tested will ensure that testing is accurate and

repeatable. This will help ensure that the Engineered Safety Feature

(ESF) ventilation systems are capable of performing their safety

function. Therefore, the proposed changes do 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 changes incorporate ASTM D3803-1989 as the testing

standard for nuclear-grade activated charcoal samples. This will

ensure that testing is accurate and repeatable. Plant structures,

systems, and components will not be operated in a different manner

as a result of these proposed changes and no physical modifications

to equipment are involved. Using the improved testing protocol does

not have the potential for creating the possibility of a new or

different type of accident from any previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed changes do not change the manner in which

structures, systems or components are operated. Revising the

standard to which activated charcoal samples are tested will ensure

that testing is accurate and repeatable. This will help ensure that

the ESF ventilation systems are capable of performing their safety

function. Therefore, the proposed changes do not involve a reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602.

NRC Section Chief: Richard P. Correia.

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

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

Date of amendment request: November 12, 1999.

Description of amendment request: The proposed change revises the

pressure-temperature limits by revising

[[Page 70082]]

the heatup, cooldown and inservice test limitations for the Reactor

Pressure Vessel to a maximum of 32 Effective Full Power Years.

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 do not modify the reactor coolant pressure

boundary, do not make changes in operating pressure, materials or

seismic loading. The proposed changes adjust the reference

temperature for the limiting beltline material to account for

radiation effects and provide the same level of protection as

previously evaluated. The proposed changes do not adversely affect

the integrity of the reactor coolant system (RCS) such that its

function in the control of radiological consequences is affected.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

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

different kind of accident previously evaluated for Quad Cities

Nuclear Power Station. No new modes of operation are introduced by

the proposed changes. The proposed changes will not create any

failure mode not bounded by previously evaluated accidents. Use of

the revised pressure-temperature (P-T) curves will continue to

provide the same level of protection as was previously reviewed and

approved.

Further, the proposed changes to the P-T curves do not affect

any activities or equipment, and are not assumed in any safety

analysis to initiate any accident sequence for Quad Cities Nuclear

Power Station. Therefore, the proposed changes do not create the

possibility of a new or different kind of accident from any

previously evaluated.

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

of safety?

The proposed changes reflect an update of the P-T curves to

extend the Reactor Pressure Vessel (RPV) operating limit to 32

Effective Full Power Years (EFPY). The revised curves are based on

the latest American Society of Mechanical Engineers (ASME) guidance

and actual operational data for the units. This proposed changes are

acceptable because the ASME guidance maintains the relative margin

of safety commensurate with that which existed at the time that the

ASME Section IX Appendix G was approved in 1974. Therefore, the

proposed changes do not involve a significant reduction in the

margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

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

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

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

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

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

Date of amendment request: November 16, 1999.

Description of amendment request: The proposed change modifies the

surveillance requirements for Functional Unit 3 on Table 4.1.A-1 due to

replacement of the Reactor Pressure Vessel Steam Dome pressure switches

with analog trip units.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the change involve a significant increase in the

probability of occurrence or consequences of an accident previously

evaluated?

During the upcoming refueling outages at Quad Cities Nuclear

Power Station, Unit 1 and Unit 2, a design change will be

implemented that upgrades the existing Reactor Vessel Steam Dome-

High instrumentation from a pressure switch to an analog trip unit

device. Analog trip units are proven technology that are more

reliable than existing equipment. Analog trip units are used in

various applications of Quad Cities Nuclear Power Station, including

the Reactor Protection System (RPS) low water level trip function.

The proposed change adds a CHANNEL CHECK and 31-day trip unit

calibration requirement for the Reactor Vessel Steam Dome Pressure--

High RPS trip function. This requirement is not applicable to the

existing instrumentation because the Barksdale pressure switches are

non-indicating and do not employ trip units.

Technical Specification (TS) requirements that govern

operability or routine testing of plant instruments are not assumed

to be initiators of any analyzed event because these instruments are

intended to prevent, detect, or mitigate accidents. Therefore, these

changes will not involve an increase in the probability of

occurrence of an accident previously evaluated. Additionally, these

changes will not increase the consequences of an accident previously

evaluated because the proposed change does not adversely impact

structures, systems, or components (SSCs). The planned instrument

upgrade is a more reliable design than existing equipment. The

proposed change establishes requirements that ensures components are

operable when necessary for the prevention or mitigation of

accidents or transients. Furthermore, there will be no change in the

types or significant increase in the amounts of any effluents

released offsite. For these reasons, the proposed changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes support a planned instrumentation upgrade

by incorporating Surveillance Requirements required to ensure

operability. The change does not adversely impact the manner in

which the instrument will operate under normal and abnormal

operating conditions. Therefore, these changes provide an equivalent

level of safety and will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The changes in methods governing normal plant operation are

consistent with the current safety analysis assumptions. Therefore,

these changes will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

The proposed change supports a planned instrumentation upgrade.

The proposed change does not affect the probability of failure or

availability of the affected instrumentation. The addition of a

CHANNEL CHECK and 31-day trip unit calibration for RPS Functional

Unit 3 (Reactor Vessel Steam Dome Pressure--High) is a conservative

change that aligns the surveillance requirements for a planned

instrumentation upgrade with that of similar instrumentation.

Therefore, it is concluded that the proposed changes will not result

in a reduction in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

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

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

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington

Date of amendment request: October 13, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.3.6.1, Table 3.3.6-1, ``Primary

Containment

[[Page 70083]]

Isolation Instrumentation.'' This amendment requests that Function 5 on

Table 3.3.6-1, ``RHR SDC System Isolation,'' be modified by removing

footnote (d). Footnote (d) states, ``Only the inboard trip system is

required in Modes 1, 2, and 3, as applicable, when the outboard valve

control is transferred to the alternate remote shutdown panel and the

outboard valve is closed.'' The outboard suction valve, RHR-V-8, is no

longer used as a high/low pressure interface in the residual heat

removal (RHR) system. Valve RHR-V-9, which is in series with valve RHR-

V-8, is now used as the high/low pressure interface valve. Valve RHR-V-

9 is operable in all modes of operation and therefore, footnote (d) is

no longer needed. The current footnote (e) will be relettered as

footnote (d) for consistency.

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 involves the probability and consequences of

accidents associated with the isolation of the RHR SDC [shutdown

cooling] mode of RHR operation. Isolation is provided if high

temperatures occur in RHR pump rooms or heat exchanger areas, if

reactor vessel water level is low, or if reactor vessel pressure is

high.

FSAR [Final Safety Analysis Report] Chapter 15, ``Accident

Analysis,'' describes two events associated with the RHR system

during SDC operation. FSAR Section 15.1.6, ``Inadvertent Residual

Heat Removal Shutdown Cooling Operation,'' describes the impact of

system operation during startup or cool-down when the reactor is

near critical. The proposed change removes the exemption for the

second trip system to isolate RHR SDC operation. There will be no

change in the probability or consequences of this accident as a

result of the proposed change.

The second accident is described in FSAR Section 15.2.9,

``Failure of Residual Heat Removal Shutdown Cooling.'' It postulates

the failure of the RHR system to function in SDC mode. The

evaluation assumes a failure of the SDC mode of operation but does

not disable the remaining modes of RHR operation. The alternate SDC

paths involve the use of the safety relief valves to establish a

cooling flow path to the containment suppression pool. That

evaluated accident does not result in any fuel failure. The proposed

change will not result in an increase in the probability of fuel

failures. The evaluated accident does result in normal coolant

activity being released to the suppression pool through the safety

relief valves. The proposed activity will not result in a change in

the release of this coolant activity. The proposed change requires

the removal of the exemption for the second trip system to isolate

SDC and will have no impact on the probability or consequences of

that accident.

Therefore, the operation of WNP-2 in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change will not cause any new inadvertent SDC

startup, loss of water inventory or loss of coolant accidents

(LOCA). New or different inadvertent RHR SDC startup accidents are

not possible because this change is only a further restriction on

system operation. The LOCA during Mode 3 is bounded by the LOCA

defined for Modes 1 and 2. No new primary system LOCA can be

initiated because of this change.

Therefore, the operation of WNP-2 in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The removal of an exemption for the second trip system, as

proposed by this change, will increase the probability that leaks

and high pressure will be isolated. Therefore, operation of WNP-2 in

accordance with the proposed amendment will not decrease the margin

of safety. Therefore, the operation of WNP-2 in accordance with the

proposed amendment 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 10 CFR 50.92(c) are

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

amendment request involves no significant hazards consideration.

Attorney for licensee: Thomas C. Poindexter, Esq., Winston &

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

NRC Section Chief: Stephen Dembek.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana

Date of amendment request: October 25, 1999.

Description of amendment request: The proposed license amendment

would revise the reactor pressure vessel (RPV) surveillance capsule

withdrawal schedule for the River Bend Station. The first surveillance

capsule would be withdrawn at 13.4 effective full power years (EFPY)

rather than 10.4 EFPY.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

Pressure-temperature (P/T) limits (RBS Technical Specifications

Figure 3.4.11-1) are imposed on the reactor coolant system to ensure

that adequate safety margins against nonductile or rapidly

propagating failure exist during normal operation, anticipated

operational occurrences, and system hydrostatic tests. The P/T

limits are related to the nil-ductility reference temperature,

RTNDT, as described in ASME [American Society of

Mechanical Engineers] Section III, Appendix G. Changes in the

fracture toughness properties of RPV beltline materials, resulting

from the neutron irradiation and the thermal environment, are

monitored by a surveillance program in compliance with the

requirements of 10 CFR [Part] 50, Appendix H. The effect of neutron

fluence on the shift in the nil-ductility reference temperature of

pressure vessel steel is predicted by methods given in RG

[Regulatory Guide] 1.99, [Revision] 2.

River Bend's current P/T limits, as well as those for the

planned increase in reactor thermal power (``Power Uprate''), were

established based on adjusted reference temperatures developed in

accordance with the procedures prescribed in RG 1.99, [Revision] 2,

Regulatory Position 1. Calculation of adjusted reference temperature

by these procedures includes a margin term to ensure conservative,

upper-bound values are used for the calculation of the P/T limits.

Revision of the first capsule withdrawal schedule will not affect

the P/T limits because they will continue to be established in

accordance with Regulatory Position 1 or other NRC [Nuclear

Regulatory Commission]-approved procedures. When permitted (two or

more credible surveillance data sets available), Regulatory Position

2 (or other NRC-approved) methods for determining adjusted reference

temperature will be followed.

This change is not related to any accidents previously

evaluated. The proposed change is a revision of the first

surveillance capsule withdrawal time, identified in TRM [Technical

Requirements Manual] Table 3.4.11-1, from 10.4 EFPY to 13.4 EFPY.

This change will not affect P/T limits as given in RBS Technical

Specifications Figure 3.4.11-1 or USAR Figures 5.3-4a and 5.3-4b.

This change will not affect any plant safety limits or limiting

conditions of operation. The proposed change will not affect reactor

pressure vessel performance as no physical changes are involved and

RBS vessel P/T limits will remain conservative in accordance with RG

1.99, [Revision] 2 requirements. The proposed change will not cause

the reactor pressure vessel or interfacing systems to be operated

outside of their design or testing limits. Also, the proposed change

will not alter any assumptions previously made in evaluating the

radiological consequences of accidents. Therefore, the probability

or

[[Page 70084]]

consequences of accidents previously evaluated will not be increased

by the proposed change.

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

or different kind of accident from any accident previously

evaluated.

The proposed change revises the first RPV material surveillance

capsule withdrawal time in TRM Table 3.4.11-1 from 10.4 EFPY to 13.4

EFPY. This proposed change does not involve a modification of the

design of plant structures, systems, or components. The proposed

change will not impact the manner in which the plant is operated as

plant operating and testing procedures will not be affected by the

change. The proposed change will not degrade the reliability of

structures, systems, or components important to safety (ITS) as

equipment protection features will not be deleted or modified,

equipment redundancy or independence will not be reduced, supporting

system performance will not be downgraded, the frequency of

operation of ITS equipment will not be increased, and increased or

more severe testing of ITS equipment will not be imposed. No new

accident types or failure modes will be introduced as a result of

the proposed change. Therefore, the proposed change does not create

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

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

As stated in Section 5.3.2 of the River Bend Safety Evaluation

Report (NUREG-0989), ``Appendices G and H of 10 CFR [Part] 50

describe the conditions that require pressure-temperature limits and

provide the general bases for these limits. These appendices

specifically require that pressure-temperature limits must provide

safety margins at least as great as those commended in the ASME Code

[American Society of Mechanical Engineers Boiler and Pressure Vessel

Code], Section III, Appendix G. * * * Until the results from the

reactor vessel surveillance program become available, the staff will

use Regulatory Guide (RG) 1.99, Revision 1 [now Revision 2], to

predict the amount of neutron irradiation damage.* * * The use of

operating limits based on these criteria--as defined by applicable

regulations, codes, and standards--will provide reasonable assurance

that nonductile or rapidly propagating failure will not occur, and

will constitute an acceptable basis for satisfying the applicable

requirements of General Design Criteria (GDC) 31.''

Bases for RBS Technical Specification 3.4.11 states: ``The P/T

limits are not derived from Design Basis Accident (DBA) analyses.

They are prescribed during normal operation to avoid encountering

pressure, temperature, and temperature rate of change conditions

that might cause undetected flaws to propagate and cause nonductile

failure of the RCPB [reactor coolant pressure boundary], a condition

that is unanalyzed. * * * Since the P/T limits are not derived from

any DBA, there are no acceptance limits related to the P/T limits.

Rather, the P/T limits are acceptance limits themselves since they

preclude operation in an unanalyzed condition.''

The proposed change will not affect any safety limits, limiting

safety system settings, or limiting conditions of operation. The

proposed change does not represent a change in initial conditions,

or in a system response time, or in any other parameter affecting

the course of an accident analysis supporting the Bases of any

Technical Specification. The proposed change does not involve

revision of the P/T limits but rather a revision of the withdrawal

time for the first surveillance capsule. The current P/T limits (and

proposed P/T limits for Power Uprate) were established based on

adjusted reference temperatures for vessel beltline materials

calculated in accordance with Regulatory Position 1 of RG 1.99,

[Revision] 2. P/T limits will continue to be revised as necessary

for changes in adjusted reference temperature due to changes in

fluence according to Regulatory Position 1 until two or more

credible surveillance data sets become available. When two or more

credible surveillance data sets become available, P/T limits will be

revised as prescribed by Regulatory Position 2 of RG 1.99,

[Revision] 2, or other NRC-approved guidance. Therefore, the

proposed changes do not involve a significant reduction in any

margins of safety.

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

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

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

amendment request involves no significant hazards consideration.

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: Robert A. Gramm.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: October 29, 1999.

Description of amendment request: The proposed license amendment

would change the River Bend Station (RBS) Updated Safety Analysis

Report (USAR), Sections 6.2 and 15.6, to incorporate a revision to the

calculation of radiological doses following a loss-of-coolant-accident

(LOCA). The LOCA dose calculation was revised as a result of (1) an

increase in the calculated positive pressure period (PPP) to account

for a new phenomenon identified in Information Notice (IN) 88-76, (2) a

more conservative Suppression Pool water volume value, (3) an

additional and more conservative liquid leakage term identified in IN

91-56, and (4) changes to the engineered safety features (ESF) systems

liquid leakage term.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated.

The analysis changes described by this proposed change to the

USAR are not initiators to events, and therefore do not involve the

probability of an accident. These modifications reflect a revision

to the post-LOCA dose calculation. USAR Section 15.6.5.1.1 states

that ``There are no realistic, identifiable events which would

result in a pipe break inside of containment of the magnitude

required to cause an accident LOCA * * * However, since such an

accident provides an upper limit estimate to the resultant effects

for this category of pipe breaks, it is evaluated without the causes

being identified.'' The analysis itself does not identify an

initiator, nor is it the initiator, of a LOCA. There was no physical

change to the plant. The increase to the positive pressure period

(PPP) was the result of inclusion of phenomena not previously

included in the analysis documented in the SAR [safety analysis

report], and does not have any impact on accident probability. The

inclusion of an NRC [Nuclear Regulatory Commission] Information

Notice (IN) 91-56 unfiltered liquid leakage term is voluntary and

conservative in nature and does not represent an additional failure

that could be construed as an initiator to the event. Therefore,

this change does not increase the probability of occurrence of an

accident evaluated previously in the safety analysis report (SAR).

This proposed change to the USAR does increase the consequences

of an accident, but the increase is not significant. While the

calculated off-site and control room doses of a LOCA did increase in

Revision 1 to the post-LOCA dose calculation (reference 1) [of

Attachment 1 to the License Amendment request, dated October 29,

1999], the dose consequences remain below the regulatory limits of

10 CFR [Part] 100 and 10 CFR [Part] 50, Appendix A, General Design

Criteria (GDC) 19 as approved per NUREG-0989 and License Amendment

98. This change first accounts for the potential effect that

differential temperature has on the PPP assumed in the off-site dose

analysis. It also conservatively includes an additional liquid

leakage term to account for concerns documented in NRC IN 91-56.

Neither of these changes has an appreciable effect on vital area

access doses. Vital area access dose calculations were not revised

since they still conservatively reflect the expected doses discussed

in USAR Section 12.3.2.4. There is no impact on equipment

qualification associated with the proposed change since other gross

conservatisms exist in those calculations (e.g., not crediting

suppression pool scrubbing) compared to the post-LOCA dose

calculations. Reanalysis of the off-site dose calculation

demonstrates that the revised doses are increased only slightly and

remain significantly less than the regulatory

[[Page 70085]]

limits. With the IN 91-56 term excluded, the increases are within

the criteria of less than 10 [percent] of the remaining margin,

which is the criteria to be applied in the revised 10 CFR 50.59 rule

for minimal increases in consequences. With the IN 91-56 term

included, only the 30 day LPZ [low-population zone] thyroid dose

exceeds the ``minimal increase'' criterion. Note the doses

documented in Table 1 [of Attachment 1 to the License Amendment

request, dated October 29, 1999], above, are less than the values

which had been documented in the SAR prior to the implementation and

NRC approval of TS [Technical Specifications] Amendment 98.

Therefore, this change does not significantly increase the

consequences of an accident previously evaluated in the SAR.

2. The proposed changes would not create the possibility of a

new or different kind of accident from any [previously] analyzed.

This change does not represent a physical change to the plant.

It does not involve initiators to any events in the SAR, nor does

the activity create the possibility for any new accidents. Rather,

this change is a result of the evaluation of the most limiting LOCA

which can occur at River Bend. Therefore, this change involves no

new system interactions and does not create the possibility of an

accident of a different type than those presently evaluated in the

SAR.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The off-site dose consequences are calculated in accordance with

regulatory guidance found in Regulatory Guide 1.3 and the SRP

[Standard Review Plan], consistent with the analyses submitted to

and approved by the NRC in support of Technical Specification

Amendment 98. It is conservatively assumed that 100 [percent] fuel

failure occurs instantaneously upon a recirculation pipe break, thus

2 of the 3 fission product barriers are immediately eliminated.

These assumptions are made without any causes for the failures being

identified. Containment is assumed to leak at its maximum allowable

leakage rate (0.26 [percent] per day) for the duration of the event.

Other leakage terms, such as engineered safety feature (ESF)

leakage, are assumed to be equal to the Technical Specification

limit. Since assumptions are made in accordance with Technical

Specification allowable values and regulatory guidance, this change

does not reduce the margin of safety as defined in the basis for any

RBS Technical Specification.

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

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

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

amendment request involves no significant hazards consideration.

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: Robert A. Gramm.

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

Station, Unit 3, St. Charles Parish, Louisiana

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

letters dated July 29, October 28, and November 11, 1999.

Description of amendment request: The amendment will revise

Technical Specification 6.9.1.11.1 by replacing the existing reference

to the Asea Brown Boveri-Combustion Engineering, Inc. (ABB CE), small

break loss-of-coolant (SBLOCA) accident emergency core cooling system

(ECCS) performance evaluation model with the revised model described in

the topical report CENPD-137, Supplement 2, P-A, April 1998.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

The SBLOCA ECCS performance evaluation is conducted to

demonstrate conformance of light water nuclear power reactors to the

ECCS acceptance criteria of 10 CFR 50.46. The proposed change is

associated with an analysis performed using the new Supplement 2

version of the ABB CE SBLOCA Model (S2M). The primary objective of

the analysis using the new model was to determine the impact of a

reduction in High Pressure Safety Injection (HPSI) pump flow rate

due to increased surveillance test measurement uncertainty. NRC

approval of the new S2M model for use in licensing applications of

CE design pressurized water reactors was obtained on December 16,

1997 (Reference 1) [of license amendment request dated July 29,

1998].

A comparison of the Waterford 3 results for the limiting SBLOCA

scenario using the new S2M model against the criteria of 10 CFR

50.46(b) is summarized below:

------------------------------------------------------------------------

Parameter Result Criterion

------------------------------------------------------------------------

Peak Cladding Temperature...... 1929 deg.F......... 2200 deg.F

Maximum Cladding Oxidation..... 8.09%.............. 17%

Core-wide Cladding Oxidation... http://www.nrc.gov (the

Electronic Reading Room).

CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz

Mill, Pennsylvania

Date of application for amendment: September 7, 1999, as

supplemented on October 1, 1999.

Brief description of amendment: This amendment reassigns the

responsibilities of the Site Manager, who works for the Westinghouse

Electric Company (a contractor to CBS), to the TR-2 Decommissioning

Project Director, who works for CBS.

Date of issuance: November 23, 1999.

Effective Date: November 23, 1999.

Amendment No: 10.

Facility License No. TR-2: This amendment changes the

decommissioning plan.

Date of initial notice in Federal Register: October 20, 1999 (64 FR

56529).

The Commission has issued a Safety Evaluation for this amendment

dated November 23, 1999.

No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket No. 50-254, Quad Cities Nuclear

Power Station, Unit 1, Rock Island County, Illinois

Date of application for amendment: March 30, 1999.

Brief description of amendment: The amendment revises the Technical

Specifications by changing Surveillance Requirement 4.6.E.2 to allow a

one-time extension of the 18-month requirement to pressure set test or

replace one half of the Main Steam Safety Valves to an interval of 24

months.

Date of issuance: November 30, 1999.

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

Amendment No.: 191.

Facility Operating License No. DPR-29: The amendment revised the

Technical Specifications.

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

24194).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 30, 1999.

No significant hazards consideration comments received: No.

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

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

Date of application for amendments: April 6, 1999.

Brief description of amendments: The amendments revised the

Technical Specifications (TS) to expand the allowable values for

Interlocks P-6 (Intermediate Range Neutron Flux) and P-10 (Power Range

Neutron Flux) in TS 3.3.1, Table 3.3.1-1, Function 16, Reactor Trip

System Interlocks, as recommended by Westinghouse.

Date of issuance: November 30, 1999.

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

within 30 days from the date of issuance.

Amendment Nos.: Unit 1-189; Unit 2-170.

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

revised the Technical Specifications.

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

27319).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 30, 1999.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio

Date of application for amendment: May 5, 1999.

Brief description of amendment: This amendment conforms the license

to reflect the transfer of Operating License NPF-58 for the Perry

Nuclear Power Plant, Unit 1, to the extent held by Duquesne Light

Company, to the Cleveland Electric Illuminating Company as previously

approved by an Order dated September 30, 1999.

Date of issuance: December 3, 1999.

Effective date: December 3, 1999.

[[Page 70095]]

Amendment No.: 108.

Facility Operating License No. NPF-58: This amendment revised the

operating license.

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

31879).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 30, 1999.

No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Plant, Units 3 and 4, Dade County, Florida

Date of application for amendments: July 27, 1999, as supplemented

October 4, 1999.

Brief description of amendments: Revises the Technical

Specifications (TS) to extend the allowed outage time, on a one-time

basis, for an inoperable emergency diesel generator from 72 hours to 7

days, to replace the Unit 3 diesel engine radiators prior to April

2000. The revision applies to Turkey Point Unit 3 only, however, Unit 4

is included administratively because the TS are combined for both

Units.

Date of issuance: November 19, 1999.

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

April 2000.

Amendment Nos.: 202 and 196.

Facility Operating License Nos. DPR-31 and DPR-41: Amendments

revised the TS.

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

46441). The supplemental letter of October 4, 1999, provided

clarification information that did not change the original no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 19, 1999.

No significant hazards consideration comments received: No.

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: October 8, 1998.

Brief description of amendments: The proposed amendments would

change the Technical Specifications for both units to place tighter

restrictions on the allowed outage time for the refueling water storage

tank water level instrumentation.

Date of issuance: November 30, 1999.

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

within 30 days.

Amendment Nos.: 232 and 215.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 31, 1999 (64 FR

47532). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 30, 1999.

No significant hazards consideration comments received: No.

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: September 10, 1999.

Brief description of amendments: The amendments revise Technical

Specification (TS) 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 revision to TS 3/4.4.7 also includes changes to Tables

3.4-1 and 4.4-3. TS Table 4.4-3 is revised to change 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

relocates 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 also corrects 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. Surveillance

Requirement 4.11.2.2 is revised to delete the phrase ``by analysis of

the Reactor Coolant System noble gases.''

Date of issuance: November 19, 1999.

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

within 45 days.

Amendment Nos.: 231 and 214.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

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

54376).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 19, 1999.

No significant hazards consideration comments received: No.

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

Unit 1, Montgomery County, Pennsylvania

Date of application for amendment: June 7, 1999.

Brief description of amendment: The amendment revised the technical

specifications (TSs) to reflect the permanent deactivation in the

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

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

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

this current plant configuration.

Date of issuance: November 18, 1999.

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

within 30 days.

Amendment No.: 138.

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

TSs.

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

54380).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 18, 1999.

No significant hazards consideration comments received: No.

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: January 15, 1999, as

supplemented January 18 and October 22, 1999.

Brief description of amendment: The amendment provides a revision

to the Technical Specifications for the FitzPatrick Nuclear Power Plant

by modifying the description of what constitutes an acceptable Local

Power Range Monitor calibration.

Date of issuance: November 22, 1999.

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

30 days.

Amendment No.: 257.

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

Technical Specifications.

Date of initial notice in Federal Register: April 10, 1999 (64 FR

11965).

The January 18, 1999, and October 22, 1999, letters provided

clarifying 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 November 22, 1999.

No significant hazards consideration comments received: No.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: June 22, 1999.

[[Page 70096]]

Brief description of amendment: This amendment changes the

Technical Specifications by extending the pressure-temperature (P-T)

limit curves to 24 effective full-power years (EFPY) and 32 EFPY. The

current P-T limit curves are valid through 16 EFPY.

Date of issuance: November 29, 1999.

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

30 days.

Amendment No.: 258.

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

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 November 29, 1999.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama

Date of amendments request: June 30 1997, as supplemented by

letters of February 22, March 19, June 30, and October 4, 1999.

Brief Description of amendments: The amendments change the

Technical Specifications (TS) to clarify surveillance requirements for

the control room emergency filtration system, penetration room

filtration system, and related storage pool ventilation system. The

changes also revised the required number of radiation monitoring

instrumentation channels, and deleted the containment purge exhaust

filter TS.

Date of issuance: November 23, 1999.

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

within 120 days from the date of issuance.

Amendment Nos.: 145 and 136.

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

the Technical Specifications.

Date of initial notice in Federal Register: September 1, 1999 (64

FR 47870).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 23, 1999.

No significant hazards consideration comments received: No.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: September 21, 1999.

Brief description of amendment: The amendment increases the

required volume of stored fuel in the diesel fuel oil storage tank as a

result of a conservative recalculation of diesel generator fuel

consumption.

Date of Issuance: November 22, 1999.

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

implemented within 30 days.

Amendment No.: 180.

Facility Operating License No. DPR-28: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 20, 1999 (64 FR

56537). The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated November 22, 1999.

No significant hazards consideration comments received: No.

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

Generating Station, Coffey County, Kansas

Date of amendment request: September 21, 1999, as supplemented by

letter dated November 5, 1999.

Brief description of amendment: The amendment extended the

effective full implementation date by six months, from December 31,

1999, to June 30, 2000, for Amendment No. 120 issued March 22, 1999,

that approved a modification to increase the storage capacity of spent

fuel assemblies at the site. The extension is due to delays fabricating

and installing the new fuel storage racks.

Date of issuance: November 30, 1999.

Effective date: November 30, 1999, to be implemented by June 30,

2000.

Amendment No.: 129.

Facility Operating License No. NPF-42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 20, 1999 (64 FR

56538). The supplemental letter of November 5, 1999, provided

additional clarifying information, did not expand the scope of the

application as originally noticed, and did not change the staff's

original proposed no significant hazards consideration determination

published in the Federal Register.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 30, 1999.

No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and

Final Determination of No Significant Hazards Consideration and

Opportunity for a Hearing (Exigent Public Announcement or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

Energy Act of 1954, as amended (the Act), and the Commission's rules

and regulations. The Commission has made appropriate findings as

required by the Act and the Commission's rules and regulations in 10

CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the

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

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

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

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

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

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

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

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

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

transcribed as appropriate and the licensee has been informed of the

public comments.

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

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

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

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

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

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

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

Commission may provide an

[[Page 70097]]

opportunity for public comment. If comments have been requested, it is

so stated. In either event, the State has been consulted by telephone

whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

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

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

significant hazards consideration is involved.

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

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

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

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

application for amendment, (2) the amendment to Facility Operating

License, and (3) the Commission's related letter, Safety Evaluation

and/or Environmental Assessment, as indicated. All of these items are

available for public inspection at the Commission's Public Document

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

electronically from the ADAMS Public Library component on the NRC Web

site, http://www.nrc.gov (the Electronic Reading Room).

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By January 14, 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 electronically from the ADAMS Public Library

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

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

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

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

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

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

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

appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

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

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

effect.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

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: November 17, 1999.

Brief description of amendments: The amendments revised the

Technical Specifications to modify the definition

[[Page 70098]]

of steam generator repair limit for axial tube imperfections detected

between the primary side surface of the tube sheet clad and the end of

the tube.

Date of Issuance: December 3, 1999.

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

within 30 days from the date of issuance.

Amendment Nos.: Unit 1-308; Unit 2-308; Unit 3-308.

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

Amendments revised the Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: Yes. The NRC published a public notice of the proposed

amendments, issued a proposed finding of no significant hazards

consideration and requested that any comments on the proposed no

significant hazards consideration be provided to the staff by the close

of business on December 2, 1999. The notice was published in the

``Greenville News,'' Greenville, SC; and the ``Anderson Independent-

Mail,'' Anderson, SC, on November 24, 1999. No comments have been

received.

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

exigent circumstances, consultation with the State of South Carolina,

and final no significant hazards consideration determination are

contained in a Safety Evaluation dated December 3, 1999.

Attorney for licensee: Richard W. Blackburn, Esquire, Winston and

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

NRC Section Chief: Richard L. Emch, Jr.

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: November 10, 1999 (PCN-510).

Brief description of amendments: The amendments modify the

Technical Specification Limiting Condition for Operation 3.4.9.b to

delete the phrase stating that two groups of pressurizer heaters be

``capable of being powered from an emergency power supply.

Date of issuance: November 22, 1999.

Effective date: November 22, 1999.

Amendment Nos.: Unit 2-161; Unit 3-152.

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

revised the Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: Yes. The NRC published a public notice of the proposed

amendments, issued a proposed finding of no significant hazards

consideration, and requested that any comments on the proposed no

significant hazards consideration be provided to the staff by close of

business November 19 , 1999. The notice was published in the ORANGE

COUNTY REGISTER on November 15-16, 1999. No public comments were

received.

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

exigent circumstances, and final determination of no significant

hazards consideration are contained in a Safety Evaluation dated

November 22, 1999.

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

California Edison Company, 2244 Walnut Grove Avenue, Rosemead,

California 91770.

NRC Section Chief: Stephen Dembek.

Dated at Rockville, Maryland, this 8th day of December 1999.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project Management, Office of Nuclear

Reactor Regulation.

[FR Doc. 99-32311 Filed 12-14-99; 8:45 am]

BILLING CODE 7590-01-P