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 Pub. L. 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 June 5, 1999, through June 18, 1999. The

last

[[Page 35200]]

biweekly notice was published on June 16, 1999 (64 FR 32284).

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Administration Services, Office of

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

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

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

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

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

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By July 30, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, Attention:

[[Page 35201]]

Rulemakings and Adjudications Staff, or may be delivered to the

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

NW., Washington DC, by the above date. A copy of the petition should

also be sent to the Office of the General Counsel, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, and to the attorney

for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-

529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.

1, 2, and 3, Maricopa County, Arizona

Date of amendments request: May 26, 1999.

Description of amendments request: The proposed amendment would

revise Technical Specification 3.3.1, ``Reactor Protective System (RPS)

Instrumentation--Operating,'' to change the RPS reactor coolant flow

trip setpoints. The change is intended to reduce spurious reactor trip

hazards.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

No. The proposed change will change the Reactor Protection

System (RPS) reactor coolant flow trip setpoints. The RPS functions

to mitigate the consequences of an accident. The changes to the low

reactor coolant flow trip setpoints will reduce or eliminate

unnecessary challenges to the RPS. Therefore, the proposed change

will not involve a significant increase in the probability of an

accident previously evaluated.

These changes will result in an increased time delay for the RPS

low reactor coolant flow trip. The reanalysis of the affected UFSAR

[updated final safety analysis report] Chapter 15 events (UFSAR

15.3.4, Reactor Coolant Pump Shaft Break with Loss of Offsite Power

and UFSAR 15.1.5, Steam System Piping Failures Inside and Outside

Containment--Modes 1 and 2 Operations), with the increased time

delay, shows that the dose consequences for these events remain

bounded by the UFSAR analysis. Therefore, this change does not

involve a significant increase in the consequences of an accident

previously evaluated.

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

different kind of accident from any accident previously evaluated?

No. The proposed change will change the RPS reactor coolant flow

trip setpoints. The RPS functions to mitigate the consequences of an

accident. The changes to the low reactor coolant flow trip setpoints

will reduce or eliminate unnecessary challenges to the RPS. The

proposed change only changes the mitigating actions of the RPS,

without changing the required function of the RPS. Therefore, the

change to the low reactor coolant flow trip setpoints does not

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

any accident previously evaluated.

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

margin of safety?

No. The proposed change will change the RPS reactor coolant flow

trip setpoints. The reanalysis of the affected UFSAR Chapter 15

events (UFSAR 15.3.4, Reactor Coolant Pump Shaft Break with Loss of

Offsite Power and UFSAR 15.1.5, Steam System Piping Failures Inside

and Outside Containment--Modes 1 and 2 Operations), with the revised

reactor coolant flow trip setpoints, shows that the minimum DNBR

[departure from nucleate boiling ratio] and SAFDLs [specified

acceptable fuel design limits] for these events remain bounded by

the UFSAR analysis. 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

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

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

amendments request involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999

NRC Section Chief: Stephen Dembek.

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: June 2, 1999.

Description of amendment request: The proposed amendment would

relocate Shearon Harris Nuclear Power Plant (HNP) Technical

Specification (TS) Section 6.5, ``Review and Audit,'' TS 6.8.2, TS

6.8.3, and TS Section 6.10, ``Record Retention,'' intact from the HNP

TS to the Quality Assurance Program Description currently located in

the HNP Final Safety Analysis Report Section 17.3. Future changes to

the associated relocated TS would be processed in accordance with 10

CFR 50.54(a). The proposed change is consistent with NUREG-1431,

Revision 1, ``Standard Technical Specifications, Westinghouse Plants,''

dated April 1995, and with the guidance provided in NRC Administrative

Letter 95-06, ``Relocation of Technical Specification Administrative

Controls related To Quality Assurance,'' dated December 12, 1995.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

This TS change relocates administrative requirements from HNP TS

to the Quality Assurance Program Description (QAPD). The proposed

amendment will not introduce any new equipment or require existing

equipment to function different from that previously evaluated in

the Final Safety Analysis Report (FSAR) or TS.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed amendment will not introduce any new equipment or

require existing equipment to function different from that

previously evaluated in the Final Safety Analysis Report (FSAR) or

TS. The changes are consistent with NUREG-1431, Revision 1 and the

Commission's Final Policy Statement on Technical Specification

improvements. The proposed amendment will not create any new

accident scenarios, because the change does not introduce any new

single failures, adverse equipment or material interactions, or

release paths.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

This TS change relocates administrative requirements from HNP TS

to the Quality

[[Page 35202]]

Assurance Program Description (QAPD). The QAPD will be revised to

include the requirements associated with this proposed change. NRC

Administrative Letter 95-06 states that administrative requirements

for review and audit and the independent safety engineering group

may be relocated from TS to the quality assurance program. HNP

proposes relocating the associated requirements from TS to the QAPD

intact. Future changes to these requirements will be processed in

accordance with 10 CFR 50.54(a). This proposed TS change is

administrative in nature and does not alter NRC acceptance limits

with respect to accident mitigation or accident analysis.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

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

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

Raleigh, North Carolina 27602

NRC Section Chief: Sheri R. Peterson.

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

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

Date of amendment request: July 22 and October 22, 1998; May 6,

1999.

Description of amendment request: The amendments would revise the

Technical Specifications (TS) to reflect the licensee's planned use of

fuel supplied by Westinghouse. The staff has published a Notice of

Consideration of Issuance of Amendments and Proposed No Significant

Hazards Consideration Determination on November 3, 1998 (63 FR 69338)

covering the July 22 and October 22, 1998, submittals. In the May 6,

1999, submittal the licensee proposed to expand the original amendment

request, revising Section 5.6.5 of the Technical Specifications.

Section 5.6.5 specifies a list of NRC-approved topical reports that the

licensee is required to use to determine reactor core operating limits.

The licensee proposed to update this list to show the current approval

status of these topical reports.

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 for the proposed changes conveyed by the May 6, 1999,

submittal. The NRC staff has reviewed the licensee's analyses against

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

below.

First Standard

No. The proposed changes to Section 5.6.5 will not affect the

safety function, and will not involve any change to the design or

operation of any plant system or component. The topical reports were

previously approved by the NRC staff under separate licensing actions.

The use of methodologies in these approved topical reports will ensure

that previously evaluated accidents remain bounding. Therefore, no

accident probabilities or consequences will be impacted.

Second Standard

No. The proposed changes would not lead to any hardware or

operating procedure change. Hence no new equipment failure modes or

accidents from those previously evaluated will be created.

Third Standard

No. Margin of safety is associated with confidence in the design

and operation of the plant; specifically, the ability of the fission

product barriers to perform their design functions during and following

an accident. The proposed changes to Section 5.6.5 do not involve any

change to plant design, operation, or analysis. Thus the margin of

safety previously analyzed and evaluated is maintained.

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

CFR 50.92(c) are satisfied for the proposed changes to Section 5.6.5.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,

422 South Church Street, Charlotte, North Carolina

NRC Section Chief: Richard L. Emch, Jr.

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

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

Date of amendment request: April 5, 1999, supplemented May 27,

1999.

Description of amendment request: The proposed amendments would

revise the Improved Technical Specifications (TS), Updated Final Safety

Analysis Report, and Core Operating Limits Report to incorporate

Topical Report (TR) DPC-NE-3005-P, ``Thermal-Hydraulic Transient

Analysis Methodology.'' This analysis has been completed for Unit 2 and

is ongoing for Units 1 and 3. Therefore, the proposed changes that

reflect the TR provisions affect Unit 2 only. Other proposed changes

affect all three units. Specifically, (1) a note to TS Surveillance

Requirement (SR) 3.4.1.2, ``RCS [Reactor Coolant System] Pressure,

Temperature, and Flow DNB [Departure from Nucleate Boiling] Limits,''

would be modified to address application of the delta-Tcold

limits; (2) TS 3.4.10, ``Pressurizer Safety Valves,'' would be modified

to increase the setpoint range of the lift settings for the pressurizer

safety valves for the Oconee unit that has been analyzed in accordance

with the TR and state that the range is not changed for the other

units; (3) a statement to SR 3.4.10.1 would be added that will specify

the pressurizer safety valve lift setpoint in order to clarify the

difference between the operability setpoint range for a test lift and

the range required when the setpoint is reset following the

surveillance test; (4) TS 3.7.4, ``Atmospheric Dump Valve (ADV) Flow

Paths,'' would be added to address the applicability and required

actions related to the ADS valves; (5) TS 3.9.7, ``Unborated Water

Source Isolation Valves,'' would be added to require valves that are

used to isolate unborated water sources to be secured in the closed

position while in Mode 6, incorporate SRs, and provide required actions

if one or more of the valves is not secured in the closed position; (6)

TS 5.6.5b would be changed to update the Core Operating Limits Report

references; and (7) the appropriate Bases would be changed to reflect

the above changes, other changes consistent with the revisions to the

TR analysis, and the Updated Final Safety Analysis Report revisions

that were provided in the submittal.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

No. The proposed changes to the Technical Specifications, Bases,

Updated Final Safety Analysis Report (UFSAR), and Core Operating

Limits Report (COLR) incorporate the accident analyses established

in Topical

[[Page 35203]]

Report DPC-NE-3005-P, ``UFSAR Chapter 15 Transient Analysis

Methodology.'' On July 30, 1997, Duke submitted Topical Report DPC-

NE-3005-P to the NRC for approval. The NRC found DPC-NE-3005-P

acceptable, with noted exceptions, in a Safety Evaluation issued on

October 1, 1998. To resolve the noted NRC exceptions, Duke submitted

Revision 1 of DPC-NE-3005-P to the NRC for review on February 1,

1999. Additional information regarding Revision 1 of DPC-NE-3005-P

was submitted on April 19 and May 5, 1999. This LAR is dependent

upon the NRC approval of Revision 1 of DPC-NE-3005-P. [This Topical

Report was approved by the NRC on May 25, 1999.]

The analyzed events are initiated by the failure of specific

plant structures, systems or components. These proposed changes do

not impact the condition or performance of those structures, systems

or components.

The revised accident analyses in DPC-NE-3005-P demonstrate that

the applicable acceptance criteria are met. In addition, the

preliminary calculations show that the applicable radiological and

environmental acceptance criteria continue to be met.

Based on the above, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any accident previously evaluated?

No. The proposed changes do not involve a physical alteration of

the plant. No new or different equipment is being installed, and no

installed equipment is being operated in a new or different manner.

Where setpoints and operating limits have been revised, the revised

accident analyses demonstrate that the applicable acceptance

criteria are met. As a result, no new failure modes are being

introduced.

Based on the above, the proposed changes do not create the

possibility of any new or different kind of accident from any

accident previously evaluated.

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

No. The margin of safety is established through the design of

the plant structures, systems and components, the parameters within

which the plant is operated, and the establishment of the setpoints

for the actuation of equipment relied upon to respond to a event.

The proposed changes do not involve a physical alteration of the

plant. No new or different equipment is being installed, and no

installed equipment is being operated in a new or different manner.

Where setpoints and operating limits have been revised, the revised

accident analyses in DPC-NE-3005-P demonstrate that the applicable

acceptance criteria are met.

Based on the above, the proposed changes do not involve a

significant reduction in a margin of safety.

Based upon the preceding evaluation, performed pursuant to 10

CFR 50.92, Duke has concluded that the proposed changes to the

Oconee Nuclear Station Technical Specifications, Bases, UFSAR, and

O2C18 COLR will not involve a significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Conee County Library, 501 West

South Broad Street, Walhalla, South Carolina

Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC.

NRC Section Chief: Richard L. Emch, Jr.

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

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

Date of amendment request: May 24, 1999

Description of amendment request: The proposed amendments would

revise the maximum local fuel pin centerline temperature safety limit

in Technical Specification 2.1.1.1 from the limit determined using the

TACO2 fuel performance computer code to the value determined using a

newer TACO3 computer code.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

The following discussion is a summary of the evaluation of the

changes contained in this proposed amendment against the 10 CFR

50.92 (c) requirements to demonstrate that all three standards for

no significant hazards consideration are satisfied. A no significant

hazards consideration is indicated if 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.

First Standard

Implementation of this amendment would not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The use of the revised maximum local fuel pin

centerline temperature limit is appropriate since the new limit uses

a fuel melt temperature which has been conservatively reduced to

account for code uncertainties in calculating fuel centerline

temperature. NRC has previously found the use of the TACO3 code by

DPC [Duke Power Company] in performing reload licensing to be

acceptable. The use of the revised limit for fuel analyzed using an

approved code ensures centerline fuel melting is avoided by ensuring

the maximum fuel temperature is less than the melting temperature of

the fuel. Therefore this change would not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Second Standard

Implementation of this amendment will not create the possibility

of a new or different kind of accident from any previously

evaluated. The use of the revised maximum local fuel pin centerline

temperature limit has no affect on accident precursors.

Implementation of this amendment will not impact any plant systems

that are accident initiators. No other modifications are being

proposed in the plant that would result in the creation of a new

accident mechanism. Also, no changes are being made to the way the

plant is operated; therefore, no new failure mechanisms will be

initiated.

Third Standard

The revised maximum local fuel pin centerline temperature limit

has been appropriately reduced to account for uncertainties in

predicting centerline fuel temperatures. NRC has previously found

the use of the TACO3 code by DPC in performing reload licensing to

be acceptable. Therefore, implementation of this amendment would not

involve a significant reduction in a margin of safety.

Therefore, Duke has concluded that the proposed amendment does

not involve a significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina

Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC.

NRC Section Chief: Richard L. Emch, Jr.

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

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

Date of amendment request: May 27, 1999.

Description of amendment request: The proposed changes would

relocate the seismic monitoring instrumentation requirements contained

in Technical Specification (TS) 3/4.3.3.3 to the Licensing Requirements

Manual based on the guidance provided in Generic Letter 95-10,

``Relocation of Selected Technical Specifications Requirements

[[Page 35204]]

Related to Instrumentation.'' The Bases section for Specification 3/

4.3.3.3 will also be relocated to the LRM. The appropriate Index pages,

Table Index page (Unit No. 1 only), TS pages and Bases pages will be

revised to reflect the removal of the seismic monitoring

instrumentation specification from the TSs. An additional specification

page will be added to reflect that Specification Number 3/4.3.3.4 is

not used. This additional page will also denote the number of the

following page. The Bases section will also be modified to denote that

Specification Number 3/4.3.3.4 is not used.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed amendment would relocate Technical Specification

(TS) 3/4.3.3.3 titled ``Seismic Instrumentation'' and the associated

Bases section to the Licensing Requirements Manual (LRM) (based on

the guidance provided in Generic Letter (GL) 95-10, ``Relocation of

Selected Technical Specification Requirements Related to

Instrumentation''). The proposed amendment would also revise the TS

Index and Beaver Valley Power Station (BVPS) Unit No. 1 List of

Tables to reflect the relocation of this TS and associated Bases.

The relocated Specification will be controlled in accordance with

the requirement of 10 CFR 50.59, ``Controls, Tests, and

Experiments.'' Additional administrative changes are also included

to reflect that Specification Number 3/4.3.3.4 is not used.

The proposed amendment does not involve a significant increase

in the probability of an accident previously evaluated because no

changes are being made to any accident initiator. No analyzed

accident scenario is being changed. The initiating condition and

assumptions remain as previously analyzed. The failure of the

seismic monitoring instrumentation to detect a seismic event is not

an accident initiating event.

The seismic monitoring instrumentation performs no role in

mitigating a seismic event or in achieving a safe shutdown condition

after a seismic event has occurred. Seismic instrumentation is not

assumed to function in the safety analysis. The seismic

instrumentation is not associated with a process variable, design

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

Design Basis Accident (DBA) or transient that either assumes the

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

product barrier. Seismic instrumentation does not actuate any

protective equipment or play any direct role in the mitigation of an

accident. The capability of the plant to withstand a seismic event

or other design basis accident is determined by the initial design

and construction of systems, structures, and components. This

instrumentation is used to alert operators to the seismic event and

evaluate the plant response.

The proposed revisions to the Index pages, Table Index page

(BVPS Unit No. 1 only), Specification pages and Bases pages are

administrative in nature and do not affect plant safety.

Therefore, the proposed amendment 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 amendment does not involve any physical changes to

the plant or the modes of plant operation defined in Appendix A of

the operating license. The proposed amendment does not involve the

addition or modification of plant equipment nor does it alter the

design or operation of plant systems. Seismic instrumentation does

not actuate any protective equipment or play any direct role in the

mitigation of an accident. The capability of the plant to withstand

a seismic event or other design basis accident is determined by the

design and construction of systems, structures, and components. This

instrumentation is used to alert operators to the seismic event and

evaluate the plant response.

Therefore, operation of the facility 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. Does the change involve a significant reduction in a margin

of safety?

The proposed amendment does not involve revisions to any safety

limits or safety system setting that would adversely impact plant

safety. The proposed amendment does not affect the ability of

systems, structures or components important to ensure the safe

shutdown of the facility, or the mitigation and control of accident

conditions within the facility. In addition, the proposed amendment

does not affect the ability of safety systems to ensure that the

facility can be maintained in a shutdown or refueling condition for

extended periods of time, or the availability of sufficient

instrumentation and control capability for monitoring and

maintaining the unit status.

The proposed revisions to the Index pages, Table Index page

(BVPS Unit No. 1 only), Specification pages and Bases pages are

administrative in nature and do not affect plant safety.

Therefore, the proposed amendment does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Section Chief: S. Singh Bajwa.

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

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

Date of amendment request: May 27, 1999.

Description of amendment request: The proposed amendments would (1)

revise the frequency for performing the CHANNEL FUNCTIONAL TEST (CFT)

of the manual initiation functional units specified in the Beaver

Valley Power Station, Unit Nos. 1 and 2, Engineered Safety Features

Actuation System (ESFAS) Instrumentation Technical Specifications (TSs)

from monthly, with an accompanying footnote which allows the manual

initiation to be tested on a refueling interval, to each refueling

interval; (2) Revise footnotes associated with TS ESFAS tables; (3)

revise associated TS Bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change revises the frequency notation specified for

the channel functional test of the manual initiation functions

listed on Table 4.3-2 of TS 3/4.3.2, ``Engineered Safety Feature

Actuation System (ESFAS) Instrumentation.'' The proposed change

revises the current TS requirement for surveillance testing these

functions to clarify that testing be performed on a refueling basis.

The revision to the surveillance frequency specified in Table 4.3-2

does not physically impact the Instrumentation, its setpoints, or

the actual frequency at which the manual initiation functions are

tested. The revision eliminates the potential for confusion

regarding the testing required for the manual initiation function by

deleting Footnote (1) to Table 4.3-2. The proposed change to the

Surveillance Requirements of Table 4.3-2 for the manual initiation

functions eliminates the need for Footnote (1). Footnote (1)

requires testing the manual actuation switches every 18 months and

performing a Channel Functional Test on all other circuitry

associated with manual safeguards actuation every 31 days. As there

is no other circuitry for which a 31 day CFT is applicable, the

proposed change simplifies the TS requirement consistent with the

current Standard TS for Westinghouse plants. Footnote (1) is

consistent with early versions of the Standard Technical

Specifications of

[[Page 35205]]

NUREG-0452; however, later versions of the Standard Technical

Specifications and the Improved Standard Technical Specifications of

NUREG-1431 simply require testing manual initiation functions on a

refueling or 18 month basis. The proposed refueling frequency for

testing this instrumentation recognizes that the manual initiation

functions can not be tested at power since this would introduce the

potential for a significant plant transient.

The deletion of Table 4.3-2 Footnote (1) resulted in renumbering

Footnote (2) to (1). In addition, expired Unit 2 Table 4.3-2

Footnote (3) (only applicable to the first refueling outage) was

also deleted. In addition, changes to the TS bases are made to

further clarify the channel functional test requirements. The

reorganization of the Table 4.3-2 footnotes and bases modifications

are considered to be editorial changes.

The manual initiation instrumentation will continue to be tested

in the same manner as before (every refueling). This test frequency

is consistent with the licensing basis for testing this

instrumentation described in the Updated Final Safety Analysis

Report (UFSAR) and with the testing frequency specified in the

standard Westinghouse Plant TS. Therefore, this test frequency is

considered adequate to verify instrumentation operability. In

addition, failure of a manual initiation function is not an accident

initiator. As such, the ESFAS instrumentation will continue to be

capable of providing the required safety functions described in the

UFSAR. Therefore, operation of the facility in accordance with the

proposed amendment does not involve a significant increase in the

probability or consequence 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?

There are no hardware changes associated with this license

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

safety-related plant system performs its safety function. No new

accident scenarios, transient precursors, failure mechanisms or

limiting single failures are introduced as a result of these

changes. These changes do not introduce any adverse effects or

challenges to any safety-related systems. No change is required to

any system configurations, plant equipment or analyses. Therefore,

these changes will not create the possibility of any new or

different kind of accident from any accident previously evaluated.

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

of safety?

The margin of safety depends on the maintenance of specific

operating parameters and systems within design requirements.

Updating the manual initiation function surveillance interval

requirements specified on ESFAS TS Table 4.3-2 and deleting Table

4.3-2 Footnote (1) reflects the standard Westinghouse Plant TS

requirements for this instrumentation and is consistent with the

design and operation of the plant as described in the UFSAR. In

addition, the proposed change does not reduce the current refueling

interval testing performed on this instrumentation. The refueling

test frequency specified for this instrumentation is consistent with

industry standards and considered adequate to ensure the affected

manual initiation functions are maintained operable. The proposed

change will improve the clarity of the TS requirement by eliminating

the potential for confusion as to when the surveillances are

required to be performed. As such, the proposed change continues to

ensure that the operation of the affected instrumentation is

maintained within its design requirements and that it continues to

be capable of providing the required safety functions described in

the UFSAR. Therefore, operation of the facility in accordance with

the proposed amendment will not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Section Chief: S. Singh Bajwa.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: June 1, 1999.

Description of amendment request: The proposed amendment would

revise the surveillance requirements and applicable Bases relevant to

inservice inspection requirements for the portions of the once-through

steam generator (OTSG) tubes adjacent to the primary cladding region of

the upper and lower OTSG tubesheets.

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:

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

Probability or Consequences of an Accident Previously Evaluated.

The OTSGs are used to remove heat from the reactor coolant

system during normal operation and during accident conditions. The

OTSG tubing forms a substantial portion of the reactor coolant

pressure boundary. An OTSG tube failure is a breach of the reactor

coolant pressure boundary and is a specific accident analyzed in the

Arkansas Nuclear One, Unit 1 (ANO-1), Safety Analysis Report (SAR).

The purpose of the periodic surveillance performed on the OTSGs

in accordance with ANO-1 Technical Specification (TS) 4.18 is to

ensure that the structural integrity of this portion of the reactor

coolant system will be maintained. The TS plugging limit of 40% of

the nominal tube wall thickness requires tubes to be repaired or

removed from service because the tube may become unserviceable prior

to the next inspection. Unserviceable is defined in the TS as the

condition of a tube if it leaks or contains a defect large enough to

affect its structural integrity in the event of an operating basis

earthquake, a loss-of-coolant accident, or a steam line or feedwater

line break. The proposed TS change allows OTSG tubes with axial TEC

[tube end cracking] indications that do not extend from the cladding

region into the carbon steel interface within the tube-to-tubesheet

rolled joint of the tubesheets to remain in service with existing

degradation exceeding the existing 40% through-wall (TW) plugging

limit.

Extensive testing and plant experience has illustrated that TEC

flaws confined to this area within the OTSG will not result in tube

burst or significant tube leakage under MSLB [main steamline break]

conditions. Potential leakage from tubes with TEC will be bounded by

the MSLB evaluation presented in the SAR. Therefore, allowing TEC

flaws in this specific region to remain in service will not alter

the conditions assumed in the current ANO-1 accident analysis for

OTSG tube failures under postulated accident conditions. In

addition, the condition of the OTSG tubes in this region are

monitored during regular inspection intervals to assess for evidence

of growth. Any growth noted will be addressed through the

operational assessment. Therefore, Entergy Operations has determined

that the identification, monitoring, assessment, and corrective

action programs * * * [associated with the proposed changes]

sufficiently support this change request.

Application of the TEC alternate repair criteria will allow

leaving tubes with TEC indications found in the defined area of the

tubesheets in service while ensuring safe operation by monitoring

and assessing the present and future conditions of the tubes.

Through the inspection, monitoring, and assessment programs

previously mentioned, and the on-line leak detection capabilities

available during plant operation, continued safe operation of ANO-1

is reasonably assured.

Therefore, the application of the TEC alternate repair criteria

* * * does not involve a significant increase in the probability or

consequences of any accident previously evaluated.

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

Different Kind of Accident from any Previously Evaluated.

The implementation of the TEC alternate repair criteria will not

result in any failure mode not previously analyzed. The OTSGs are

passive components. The intent of the TS surveillance requirements

are being met by these proposed changes in that adequate structural

integrity will be maintained. Potential leakage under MSLB

conditions will remain bounded by the current SAR analysis.

Additionally, the proposed change does not introduce any new modes

of plant operation.

[[Page 35206]]

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

or different kind of accident from any previously evaluated.

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

Margin of Safety.

The application of an alternate repair criteria for TEC provides

adequate assurance with margin that ANO-1 steam generator tubes will

retain their structural integrity under normal and accident

conditions. The structural requirements of TEC affected tubes have

been evaluated satisfactorily and meet or exceed regulatory

requirements. The tubing region where TEC occurs is constrained

within the tubesheet bore; therefore, there is no additional risk

associated with tube rupture. Main steam line break leakage rates

for these tubes are reasonably assured to remain within the

assumptions of the accident analysis by proper application of the

TEC alternate repair criteria program. Because no appreciable impact

is evidenced on the tubes structural integrity or its potential

leakage rate, the margin to safety remains unaltered.

Therefore, this change does not involve a significant reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

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

NRC Section Chief: Robert A. Gramm.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,

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

Date of amendment request: June 1, 1999.

Description of amendment request: The amendments would revise the

St. Lucie, Units 1 and 2, Technical Specifications (TS), Sections

3.5.2, to allow up to 7 days to restore an inoperable Low Pressure

Safety Injection System train to operable status. The amendments would

also revise the associated surveillance requirements and TS Bases

sections to be consistent with the revisions to TS Section 3.5.2. Minor

editorial changes for the specified Recirculation Actuation Signal

(RAS) verification test are also included to ensure the terminology

used in the specification is consistent with plant design.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments for St. Lucie Plant, Units 1 and 2 will

extend the action completion/allowed outage time (AOT) for a single

inoperable Low Pressure Safety Injection (LPSI) train from 72 hours

to 7 days. A LPSI train is designed as a part of each Emergency Core

Cooling System (ECCS) subsystem to supplement Safety Injection Tank

(SIT) inventory during the early stages of mitigating a Design Basis

Accident. As such, components of the LPSI system are not accident

initiators, and an extended AOT to restore operability of an

inoperable LPSI train would not increase the probability of

occurrence of accidents previously analyzed.

The safety analyses for both St. Lucie Units demonstrate that

ECCS performance acceptance criteria are satisfied with only one of

the two redundant ECCS subsystems operating during the postulated

Design Basis Accident. The proposed technical specification

revisions involve the AOT for a single inoperable LPSI train, and do

not change the conditions assumed for the minimum amount of

operating equipment needed for accident mitigation. Therefore, the

consequences of an accident previously evaluated will not be

significantly increased.

In addition to the preceding evaluation, a Probabilistic Safety

Analysis (PSA) was performed to quantitatively assess the risk

impact of the proposed amendments. It was concluded from the results

of that assessment that the risk contribution of the AOT extension

is very small, and that the net impact of the proposed amendment can

be risk beneficial.

The editorial corrections proposed for the specified RAS

verification test do not alter existing test requirements and have

no impact on the accident analyses. Therefore, operation of either

facility in accordance with its proposed amendment would not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendments will not change the physical plant or

the modes of plant operation defined in either Facility License. The

changes do not involve the addition or modification of equipment nor

do they alter the design of plant systems. Therefore, operation of

either facility in accordance with its proposed amendment would not

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

any accident previously evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

The margin of safety associated with the ECCS system is

established by acceptance criteria for system performance defined in

10 CFR 50.46. The proposed amendments will not change these

acceptance criteria or the operability requirements for equipment

that is used to achieve such performance as demonstrated in the

plant safety analyses. Moreover, an integrated assessment of the

risk impact of extending the AOT for a single inoperable LPSI train

has concluded that the risk contribution is very small, LPSI system

reliability can potentially be improved, and the net impact of the

proposed change can be risk beneficial. The editorial corrections

proposed for the specified RAS verification test do not alter

existing test requirements and have no impact on the accident

analyses. Therefore, operation of either facility in accordance with

its proposed amendment would not involve a significant reduction in

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

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

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

NRC Section Chief: Sheri R. Peterson.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: May 13, 1999.

Description of amendment request: The proposed amendment would make

changes to the TMI-1 Facility Operating License No. DPR-50 Sections

2.a, 2.c.(3), and 2.c.(7) to delete obsolete or outdated portions of

the license conditions, and would change the Bases for Technical

Specification 3.1.1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or consequences of an accident previously

evaluated. Most of the proposed amendment is only administrative; it

adds to the Technical Specifications generic references to various

documents. These changes have no affect upon the plant design or

operation.

[[Page 35207]]

The proposed change to the Technical Specification Bases 3.1.1

is the removal of the specified pressurizer code safety valve flow-

rate for which no basis could be found and the acceptance of a 3%

setpoint drift (as-found) as per the ASME code. The 3% code limit is

in accordance with the plant's Inservice Test Program submittal,

which was evaluated by the NRC staff for the current 10 year

interval and documented under NRC TAC No. M93777. The [c]orrect

pressurizer code safety valve flow is provided in the FSAR Table

4.2-8. The proposed change is supported by a revise[d] Startup

Accident analysis with the revised safety valve flow-rate at the 3%

setpoint drift, which demonstrated that the acceptance criteria for

the event were met with considerable margin. The proposed change

does not affect the Technical Specification 3.1.1.a, pressurizer

code safety valve operable (as-left) requirement of [plus or minus]

1%.

Therefore, operation in accordance with the proposed amendment

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated, because no

new failure modes are created by the proposed changes. The

administrative changes are cosmetic and have no impact on plant

design or operation.

3. Operation of the facility in accordance with the proposed

amendment will not involve a significant reduction in a margin of

safety. The proposed amendment does not change any operating limits

for reactor operation.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Section Chief: S. Singh Bajwa.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: May 26, 1999.

Description of amendment request: The proposed amendment would

approve changes to the TMI-1 Updated Final Safety Analysis Report

(UFSAR) which would allow use of the EPRI (Electric Power Research

Institute) Conservative Deterministic Failure Margin (CDFM) methodology

for seismic analysis of the portions of the auxiliary steam line

located in the Auxiliary, Control and Fuel Handling buildings at TMI-1.

The licensee determined that these changes to the UFSAR required prior

NRC approval in accordance with 10 CFR 50.59.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment, use of CDFM methodology for the

analysis of the auxiliary steam system piping, would not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The analysis of the auxiliary steam pipe using the CDFM

methodology demonstrates that the pipe wall will maintain integrity

sufficient to prevent adverse impact on safety related equipment

during a safe shutdown earthquake (SSE). The methodology is based on

actual earthquake experience data and has been shown to be adequate

to demonstrate that piping systems will maintain integrity. The CDFM

methodology was developed by experts in the field of seismic

analysis and is based on actual earthquake experience and the

results of dynamic tests with large seismic accelerations. The

methodology provides a conservative mechanism for analytically

predicting performance during actual earthquakes, and thus its

application would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. The proposed amendment, use of CDFM methodology for the

analysis of the auxiliary steam system piping, would not create the

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

previously evaluateed.

No changes to plant systems, structures or components are

proposed and no changes to methods of operation [of the plant] are

involved.

3. The proposed amendment, use of CDFM methodology for the

analysis of the auxiliary steam system piping, would not involve a

significant reduction in a margin of safety.

No changes are proposed to operating limits or safety system

settings, or to accident analysis acceptance criteria. The CDFM

methodology provides a conservative mechanism for analytically

predicting system performance during actual earthquakes. Its

application to the auxiliary steam system piping would not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Section Chief: S. Singh Bajwa.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: June 4, 1999.

Description of amendment request: The amendment revises decay heat

removal capability requirements to ensure that at least two active

methods of decay heat removal capability will be available during

shutdown conditions except when the reactor vessel head is removed and

the fuel transfer canal water level is greater than or equal to 23 feet

above the reactor vessel flange.

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:

GPU Nuclear has determined that this Technical Specification

Change Request poses no significant hazards as defined by NRC in 10

CFR 50.92. 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 because the

proposed changes would remove exceptions for decay heat removal

system operability requirements during the time the plant is in a

Refueling Shutdown with the RCS loop not filled. The proposed

changes effectively add requirements to maintain redundancy in decay

heat removal systems.

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

from any accident previously evaluated because the proposed changes

would not introduce any new failure modes or modify existing

systems.

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

the proposed amendment would not involve changes to the safety

limits, limiting safety system settings, or operating limits.

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

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

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

amendment request involves no significant hazards consideration.

[[Page 35208]]

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Section Chief: S. Singh Bajwa.

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

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

Date of amendment request: April 19, 1999.

Description of amendment request: The proposed amendment would

replace the current set of technical specifications for the Millstone

Unit 1 plant with a new set of technical specifications for the

permanently shutdown status of the plant.

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, a summary of which is presented below:

The proposed change does not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

This proposed change is consistent with the STS [standard

technical specifications]. The relocation of requirements from the

MP1 TS [Millstone Unit 1 Technical Specifications] to the licensee

controlled documents is consistent with the criteria set forth in 10

CFR 50.36 for the content of Technical Specifications. The removal

of definitions, generic LCO [limiting condition for operation]

actions and generic surveillance requirements has no impact on

facility SSCs [structure, system, and components] or the methods of

operation of such SSCs. The deletion of design features and safety

limits not applicable to the permanently shutdown and defueled

status of MP1 has no impact on the remaining DBA [design-basis

accident], the fuel handling accidents in the fuel storage pool. The

removal of LCOs and surveillance requirements which are related only

to the operation of the nuclear reactor or only to the prevention,

diagnosis or mitigation of reactor-related transients or accidents

do not affect the applicable DBA previously evaluated. The critical

safety functions involving core reactivity control, reactor heat

removal, reactor coolant system inventory control and containment

integrity are no longer necessary at MP1. The proposed accidents

involving damage to the reactor coolant system, main steam lines,

reactor core, and the subsequent release of radioactive material are

no longer possible at MP1. Fuel pool cooling and makeup related

equipment and support equipment (e.g., electrical power systems) are

not required to be continuously available since recent analysis

demonstrated that there is up to ten days before fuel storage pool

boiling to effect repairs, establish alternate sources of make up

flow, or establish steady state natural air circulation cooling of

the Reactor Building atmosphere and fuel storage pool water in the

event of a loss of cooling and makeup flow to the fuel pool. The

radioactive decay of the irradiated fuel since shutdown of the

reactor in November, 1995 has reduced the consequences of the fuel

handling accident to levels well below those previously analyzed.

The relevant parameter (water level) associated with the fuel pool

provides an initial condition for the fuel handling accident

analyses and is included in the PDTS [Permanently Defueled Technical

Specifications]. The Reactor Building crane LCOs are retained to

preserve the engineered controls which preclude a spent fuel cask

drop from occurring over the fuel storage pool. The deletion and

modification of provisions of the administrative controls do not

directly affect the design of SSCs necessary for safe storage of

irradiated fuel or the methods used for handling and storage of such

fuel in the fuel pool. The relocation of administrative controls

related to quality assurance to the Northeast Utilities Quality

Assurance Program is also consistent with the guidance provided in

NRC Administrative Letter AL 95-06, ``Relocation of Technical

Specification Administrative Controls Related to Quality

Assurance,'' dated December 12, 1995. The changes to the

administrative controls are administrative in nature and do not

affect any accidents applicable to the safe storage of irradiated

fuel or the permanently shutdown and defueled condition of the

reactor. Therefore, the proposed changes to the MP1 TS do not

involve any increase in the probability or consequences of any

accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes have no impact on facility SSCs affecting

the safe storage of irradiated fuel or on the methods of operation

of such SSCs, or handling and storage of such fuel. These changes

are consistent with the STS and add to the clarity and ease of use

of the proposed PDTS. The removal of Technical Specifications which

are related only to the operation of the nuclear reactor or only to

the prevention, diagnosis, or mitigation of reactor-related

transients or accidents cannot result in different or more adverse

failure modes or accidents than previously evaluated because the

reactor is permanently shutdown and defueled and MP1 is no longer

authorized to operate the plant. The proposed deletion of provisions

of the MP1 TS do not affect systems credited in the accident

analyses for the fuel handling accident in the fuel storage pool at

MP1. The proposed PDTS continue to require proper control and

monitoring of safety significant parameters and activities. The

proposed restriction on the fuel pool level is fulfilled by normal

operating conditions and preserves initial conditions assumed in the

analyses of the postulated DBA. Reactor Building crane LCOs are

retained from current Technical Specifications to preclude the

possibility of a spent fuel cask drop over the fuel storage pool.

Therefore, the proposed changes to this section of the MP1 TS would

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

from any accident previously evaluated.

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

The deletion of provisions of the MP1 TS, which are not related

to the storage of irradiated fuel or which are inconsistent with the

scope of the STS, will not affect the analyses of the remaining DBA

applicable to MP1. The postulated DBAs involving the reactor are no

longer possible due to the permanently shutdown and defueled

condition of the reactor. The requirements for SSCs which have been

deleted from the MP1 TS are not credited in the existing accident

analyses for the remaining applicable postulated accidents and

therefore, do not contribute to the margin of safety associated with

the accident analysis. Therefore, the proposed changes to this

section of the MP1 TS do not involve any reduction in a margin of

safety.

Conclusion

NNECO has concluded that the proposed change to the MP1 Technical

Specifications does not involve a significant hazards consideration as

defined by 10 CFR 50.92.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

Juliano, 49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Section Chief: Michael T. Masnik.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: January 25, 1996, as supplemented April

26, 1996, September 12, 1996, March 17, 1997, September 9, 1997,

December 30, 1998, and May 19, 1999.

Description of amendment request: The proposed changes extend the

allowed outage time for an emergency diesel generator (EDG) system from

7 to

[[Page 35209]]

14 days. At FitzPatrick, an EDG system consists of 2 EDGs powering one

of two emergency AC power buses. The proposal includes provisions for a

Configuration Risk Management Program (CRMP) consistent with the

guidance of Regulatory Guide (RG) 1.177, ``An Approach for Plant-

Specific, Risk-Informed Decisionmaking: Technical Specifications.'' The

NRC staff had previously published a notice on these topics on March

27, 1996 (61 FR 13532). This revised notice on these topics is required

to address revisions made in the licensee's supplemental submittals.

The licensee's January 25, 1996, submittal also proposed two line-

item changes to reduce EDG testing at power and to revise AC power

requirements for cold shutdown and refueling modes. The two line-item

changes have not been affected by the supplemental information provided

by the licensee, so the March 27, 1996, proposed finding of no

significant hazards considerations remains valid for these items.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

additional changes to the proposed Amendment discussed above, would not

involve a significant hazards consideration as defined in 10 CFR 50.92,

since it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to the Technical Specifications will allow

longer Allowed Out of Service Times to perform necessary repair and

maintenance on Emergency Diesel Generators while at power. This

extended AOT [allowed outage time] will enhance scheduling of

preventive maintenance of individual EDGs without significantly

increasing the probability or consequences of an accident previously

evaluated. The risk evaluations for the EDGs determined that the

probability of an accident by increasing the AOT for an EDG System

from 7 days to 14 days is non-risk-significant.

Increasing the EDG AOT does not involve physical alteration of

any plant equipment and does not affect analysis assumptions

regarding functioning of required equipment designed to mitigate the

consequences of accidents. Further, the severity of postulated

accidents and resulting radiological effluent releases will not be

affected by the increased AOT for an EDG System.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

[The CRMP provides administrative controls to ensure equipment

configurations do not result in any significant increase in plant

risk. In RG 1.177, the NRC staff established a standard for the

content of the CRMP. The licensee's proposal is consistent with that

standard, and so does not involve a significant increase in the

probability or consequences of an accident previously evaluated.]

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

from any accident previously evaluated.

Extending the AOT for an EDG system does not necessitate

physical alteration of the plant or changes in parameters governing

normal plant operation. Thus, this change does not create the

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

previously evaluated for [the] JAF [FitzPatrick] plant.

[The CRMP provides administrative controls to ensure equipment

configurations do not result in any significant increase in plant

risk. These administrative controls do not create any new equipment

configurations, or provide for operation of equipment in a new or

different manner. Therefore, the CRMP does not create the

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

previously evaluated.]

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

As discussed above, a Fitzpatrick evaluation determined that the

change in risk associated with extending the AOT for a[n] EDG System

is non-risk-significant. In addition, the design provides adequate

redundancy for safe shut down during the AOT with an EDG System out

of service. This is supported by the LOCA [loss-of-coolant accident]

analyses including analyses for long term suppression pool cooling

and reactor shutdown cooling.

[The CRMP provides administrative controls to ensure equipment

configurations do not result in any significant increase in plant

risk. These administrative controls do not create any new equipment

configurations, or provide for operation of equipment in a new or

different manner. Therefore, the proposed CRMP does not involve a

significant reduction in the margin of safety.]

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek

Generating Station, Salem County, New Jersey

Date of amendment request: May 24, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to correct typographical and

editorial errors, and is considered administrative in nature.

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.

The proposed editorial and administrative changes involve

typographical errors and/or reflect changes that were previously

reviewed and approved by the NRC. These changes, therefore, do not

modify or add any initiating parameters that would significantly

increase the probability or consequences of any previously analyzed

accident.

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

or different kind of accident from any accident previously

evaluated.

These proposed changes do not involve any potential initiating

events that would create the possibility of a new or different kind

of accident. Therefore, the proposed changes do not create the

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

previously evaluated.

(3) The proposed change does not involve a significant reduction

in a margin of safety.

These changes are editorial in nature and/or reflect information

previously reviewed and approved by the NRC. The proposed changes

will make the information in the TS consistent with that already

approved by the NRC. Therefore, the proposed changes do not involve

a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

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

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

NRC Section Chief: James W. Clifford.

[[Page 35210]]

Sacramento Municipal Utility District (the District), Docket No. 50-

312, Rancho Seco Nuclear Station, Sacramento County, California

Date of amendment request: April 23, 1999.

Description of amendment request: The proposed amendment would

change Permanently Defueled Technical Specification (PDTS) D3/4.1,

``Spent Fuel Pool Level,'' to replace a specific reference to spent

fuel pool (SFP) level alarm switches with a generic reference to SFP

level instrumentation. This would allow the licensee to replace the old

level alarm switches with a new ultrasonic level transmitter.

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:

PA-193 will not create a significant increase in the probability

or consequences of an accident previously evaluated in the SAR

[Safety Analysis Report], because the proposed PDTS change is

editorial in nature and only changes the type of equipment that is

referenced in surveillance specification D4.1.2. The SFP level

instrument reference in D4.1.2 is changed from a specific reference

(i.e., SFP level alarm switches) to a more generic reference (i.e.,

SFP level instrumentation). In addition:

1. SFP level monitoring instrumentation is not relied on to

mitigate the consequences of the accidents analyzed in the SAR

(i.e., Fuel Handling Accident, Loss-Of-Offsite-Power event, Liquid

Tank Ruptures, and Decommissioning Accidents),

2. PA-193 does not alter the SFP level monitoring, SFP cooling,

or fuel handling functions during the PDM [Permanently Defueled

Mode],

3. PA-193 continues to require an 18-month calibration of SFP

level instrumentation, and

4. SFP level and alarm indication in the Control Room is

maintained with the new SFP level instrumentation. Also, the SFP

level alarm setpoints remain unchanged with the new SFP level

detection system.

PA-193 will not create the possibility of a new or different

type of accident than previously evaluated in the SAR, because SFP

level instrumentation does not provide any control function and does

not affect any equipment associated with SFP cooling, fuel handling,

or inventory control. The proposed wording change to PDTS D4.1.2

accommodates upgrading the SFP level instrumentation without

changing the intent of surveillance specification D4.1.2. Also, the

new SFP level detection system will (1) maintain the existing SFP

level alarm setpoints and Control Room indication features and (2)

have no adverse impact on the SFP level monitoring function.

PA-193 will not involve a significant reduction in the margin of

safety, because the proposed PDTS change is editorial in nature and

necessary and only accommodates replacing an unreliable, antiquated

SFP level monitoring system with a new, state-of-the-art, ultrasonic

level detection system. The new SFP level detection system will

improve the accuracy, reliability, and serviceability of the SFP

level monitoring function. The District is maintaining the

requirement to perform a[n] SFP level calibration and is only

changing the type of equipment that is referenced in D4.1.2 from a

specific reference (i.e., SFP level alarm switches) to a more

generic reference (i.e., SFP level instrumentation).

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 amendment involves no significant hazards consideration.

Local Public Document Room location: Central Library, Government

Documents, 828 I Street, Sacramento, California 95814.

Attorney for licensee: Dana Appling, Esq., Sacramento Municipal

Utility District, P.O. Box 15830, Sacramento, California 95852-1830.

NRC Section Chief: Michael T. Masnik.

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

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

County, California

Date of amendment requests: June 8, 1999 (PCN-495).

Description of amendment requests: The licensee has re-evaluated

its small break loss-of-coolant accident (SBLOCA) using ABB Combustion

Engineering (ABB-CE) S2M evaluation model. Based on this re-evaluation,

the licensee proposes to revise the Technical Specifications (TSs) for

the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 to

reflect that charging flow is not required to mitigate the effects of

the SBLOCA, add a surveillance requirement to verify that each charging

pump is operable for boration based on the Inservice Testing Program,

increase the maximum as-found lift pressure positive tolerance of main

steam safety valves (MSSVs) from +1% to +2% of the lift setting, and

list the ABB-CE S2M model as an acceptable method for determining

linear heat rate. The licensee will also revise the TS Bases and the

Updated Final Safety Analysis Report (UFSAR) to reflect the proposed

changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of any accident previously evaluated?

Response: No.

The new Small Break Loss Of Coolant Accident (SBLOCA) evaluation

model (ABB Combustion Engineering (ABB-CE) S2M SBLOCA evaluation

model, CENPD 137 Supplement 2-P-A, ``Calculative Methods of the ABB-

CE Small Break LOCA Evaluation Model,'' dated April 1998) more

accurately models the heat transfer mechanisms that occur during a

SBLOCA. As a result of this modeling improvement, there is no longer

a need to credit charging flow during a SBLOCA. The reanalysis, with

an as-found tolerance of +2%/-3% of the lift setting on Main Steam

Safety Valves (MSSVs) 2(3)-PSV-8401 and 2(3)-PSV-8410 in Table

3.7.1-2, determined that the peak cladding temperature (PCT) that

occurs in a SBLOCA is within the acceptance criteria limit of 2200

[degrees] F specified in 10CFR50.46.

This proposed change removes the charging pump Emergency Core

Cooling System (ECCS) surveillance requirement from the Technical

Specifications (TS) which effectively removes the charging system

from the ECCS. This is based on the SBLOCA reanalysis using the new

ABB-CE S2M SBLOCA evaluation model. The reanalysis using the new

model did not credit charging system flow to the reactor coolant

system.

Because this proposed change to remove the charging pump ECCS

flow surveillance requirement is based on a reanalysis of the SBLOCA

rather than physical changes to the plant or the way it is operated,

the probability of the SBLOCA is not affected. The results of the

reanalysis demonstrate the consequences of the SBLOCA without

charging flow do not exceed the consequences of the limiting LOCA.

This is based on the fact that the SBLOCA PCT [peak clad

temperature] does not exceed the limiting large break LOCA PCT.

The addition of Surveillance Requirement (SR) 3.1.9.5 to require

the charging pump to be tested in accordance with the Inservice

Testing (IST) program will ensure that the charging pumps remain

capable of performing their emergency boration requirements.

Use of the NRC approved ABB-CE S2M SBLOCA analysis methodology

identified in TS 5.7.1.5 for calculating the core operating limits

further assures that there is no significant increase in the

probability or consequences of any accident.

Therefore, the probability or consequences of any accident

previously evaluated are not increased.

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

accident from any previously evaluated?

Response: No.

This change does not involve a physical change to the plant, or

a change to the way the plant is operated. The as-left tolerance of

[plus or minus] 1% on MSSVs 2(3)-PSV-8401 and 2(3)-PSV-8410 in Table

3.7.1-2 is not being changed. The charging system will still be

verified capable of meeting its emergency boration requirements.

[[Page 35211]]

Use of the NRC approved ABB-CE S2M SBLOCA analysis methodology

identified in TS 5.7.1.5 for calculating the core operating limits

further assures that there is no increase in the possibility of a

new or different kind of accident from any previously evaluated.

Therefore, the possibility of a new or different kind of accident

from any previously evaluated is not created.

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

Response: No.

This proposed change to remove the ECCS surveillance requirement

for the charging pumps, and increase the as-found tolerance on MSSVs

2(3)-PSV-8401 and 2(3)-PSV-8410, is based on a SBLOCA reanalysis

using the new ABB-CE S2M SBLOCA evaluation model. The NRC Safety

Evaluation for the ABB-CE S2M evaluation model determined that the

new evaluation model contains sufficient conservatism such that an

adequate margin of safety exists when the S21VI evaluation model is

used. The results of the SBLOCA reanalysis are within the acceptance

criteria specified in 10 CFR 50.46.

Testing of the charging pumps per the Inservice Testing Program,

combined with the existing Technical Specification 3.1.9--``Boration

System--Operating'' surveillance requirements ensure that the

emergency boration requirements remain met without any reduction in

a margin of safety.

Use of the NRC approved S2M ABB-CE SBLOCA analysis methodology

identified in TS 5.7.1.5 for calculating the core operating limits

further assures that there is no significant reduction in any margin

of safety.

Therefore, a significant reduction in margin of safety is not

involved.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

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

California Edison Company, 2244 Walnut Grove Avenue, Rosemead,

California 91770.

NRC Section Chief: Stephen Dembek.

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

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

Date of amendment request: June 7, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 2.2.1, Reactor Trip System (RTS)

Instrumentation Setpoints, and TS 3.3.2, Engineered Safety Features

Actuation System (ESFAS) Instrumentation, and the associated Bases, by

removing the Total Allowance (TA), Sensor Error (S), and Z terms from

the RTS and ESFAS Instrumentation Trip Setpoints Tables. This would

replace the five-column methodology with a two-column methodology that

consists of the trip setpoint and allowable value columns.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed change eliminates the option to evaluate the

equation (Z+R+S [is less than or equal to] TA), within 12 hours,

from Technical Specification 2.2.1, when the trip setpoint is

outside the allowable value limit. The equation established a

threshold for submitting a Licensee Event Report. The change does

not affect the probability of an accident. The evaluation of the

equation is an administrative provision and has no relevance to the

initiation of any analyzed event. The consequences of an accident

are not affected. The change will not alter assumptions relative to

the mitigation of an accident or transient event.

The proposed amendment is a programmatic and administrative

change that does not physically alter safety-related systems, nor

does it affect the way in which safety-related systems perform their

functions. Because the design of the facility and system operating

parameters are not being changed, the proposed amendment does not

involve an increase in the probability or consequences of any

accident previously evaluated.

The proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed amendment is a programmatic and administrative

change that does not physically alter safety-related systems, nor

does it affect the way in which safety-related systems perform their

functions. The changes in methods governing normal plant operation

are consistent with current safety analysis assumptions. The

proposed change eliminates the option to evaluate the equation

(described above) within 12 hours, when the trip setpoint is outside

the allowable limit. Because the design of the facility and system

operating parameters are not being changed, the proposed amendment

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

accident from any previously evaluated.

The proposed change does not involve a significant reduction in

a margin of safety.

The proposed amendment is a programmatic and administrative

change that provides assurance that plant operations continue to be

conducted in a safe manner. As stated above, the proposed amendment

does not physically alter safety-related systems, nor does it affect

the way in which safety-related systems perform their functions. The

proposed change eliminates the option to evaluate the equation

(described above) within 12 hours, when the trip setpoint is outside

the allowable limit.

The margin of safety is not affected by eliminating an

administrative provision in Technical Specifications. The

determination for submitting a Licensee Event Report when a trip

setpoint is outside the allowable value will be performed with the

guidelines of 10CFR50.73. The safety analysis assumptions will still

be maintained, thus, no question of safety exists. Because the

design of the facility and system operating parameters are not being

changed, the proposed amendment does not involve a significant

reduction in the margin of safety.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas

77488.

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

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

NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority, Docket Nos. 50-260, 50-296, Browns Ferry

Nuclear Power Plant, Units 2 and 3. Limestone County, Alabama

Date of amendment request: March 12, 1997 as supplemented by

letters dated March 30, 1999, April 23, 1999 and June 18, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specifications to extend, from 7 days to 14 days,

the Allowable Outage Time (AOT) applicable to an inoperable emergency

diesel 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:

No Significant Hazards Consideration Determination

TVA has concluded that operation of BFN in accordance with the

proposed change to the TS does not involve a significant hazards

consideration. TVA's conclusion is based on it's evaluation, in

accordance with 10 CFR 50.91(a)(1), of the three standards set forth

in 10 CFR 50.92(c).

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The EDGs are designed as backup AC power sources in the event of

loss of off-site

[[Page 35212]]

power. The proposed AOT does not change the conditions, operating

configurations, or minimum amount of operating equipment assumed in

the safety analysis for accident mitigation. No changes are proposed

in the manner in which the EDGs provide plant protection or which

create new modes of plant operation. In addition, a PSA evaluation

concluded that the risk contribution of the AOT extension is non-

risk significant. Therefore, the proposed amendment does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change does not introduce any new modes of plant

operation or make physical changes to plant systems. Therefore,

extension of the allowable AOT for EDGs does not create the

possibility of a new or different accident.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

BFN's emergency AC system is designed with sufficient redundancy

such that an EDG may be removed from service for maintenance or

testing. The remaining EDGs are capable of carrying sufficient

electrical loads to satisfy the UFSAR requirements for accident

mitigation or unit safe shutdown.

Increasing the allowable EDG AOT will likely increase EDG

unavailability on the average since it expected that the provision

would occasionally be used to accommodate unplanned major EDG

maintenance. However, a conservative PSA evaluation concluded that

the risk contribution of the AOT extension is non-risk significant.

For the 12-year EDG PM work activity, it is expected that the

proposed TS would actually reduce unavailability since multiple

outages would not be necessary to accomplish the maintenance

activity.

The proposed change does not impact the redundancy or

availability requirements of off-site power supplies or change the

ability of the plant to cope with station blackout events. For these

reasons, the proposed amendment does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Athens Public Library, 405 E.

South Street, Athens, Alabama.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Section Chief: Sheri R. Peterson.

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

Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

Date of amendment request: May 4, 1999, as supplemented by letter

dated June 4, 1999.

Brief description of amendments: The proposed license amendments

would revise the Technical Specifications for CPSES, Units 1 and 2.

Specifically, the changes would revise the surveillance requirements

associated with the plant battery and emergency diesel generators, and

correct miscellaneous editorial errors that resulted from the issuance

of Amendment No. 64. The original application was noticed and published

in the Federal Register on June 2, 1999 (64 FR 29715). The June 4,

1999, supplement provided proposed additional editorial corrections.

The supplemental information is being noticed herein to address the

issue of no significant hazards consideration.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

probability or consequence of an accident previously evaluated?

(1) Batteries are used to support mitigation of the consequences

of an accident, and are not considered to be an initiator of any

previously analyzed accident. The proposed change would not effect

the design or performance of the batteries. The allowance to perform

the modified performance discharge test in lieu of the service test

at any time is permissible since the test's discharge rate envelopes

the duty cycle of the service test. Therefore, the allowance for

unrestricted substitution of the modified performance discharge test

in lieu of the service discharge test does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

(2) The diesel generators are used to support mitigation of the

consequences of an accident, and are not considered to be an

initiator of any previously analyzed accident. The proposed change

does not affect the accident analysis assumption that the DG reaches

minimum conditions to accept load within 10 seconds. The ability of

the DG to maintain steady state operation within 10 seconds is not

an accident analysis assumption and is primarily used to identify

degradation of governor and voltage regulator performance.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

(3) The editorial changes are non-technical and therefore do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

different kind of accident from any accident previously evaluated?

(1) The allowance for unrestricted substitution of the modified

performance discharge test in lieu of the service discharge test

does not involve any physical alteration to the plant. No new

failure mechanisms will be introduced and the change does not affect

the ability of the batteries to fulfill their safety-related

function. Therefore, this change does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

(2) The separation of the DG start surveillance criteria into

those criteria required to be met within 10 seconds, and those

criteria required to be met following achievement of steady state

conditions, does not involve any physical alteration to the plant.

No new failure mechanisms will be introduced and the change does not

affect the ability of the DGs to fulfill their safety-related

function. Therefore, this change does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

(3) The editorial changes are non-technical and therefore do not

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

any accident previously evaluated.

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

margin of safety?

(1) The allowance for unrestricted substitution of the modified

performance discharge test in lieu of the service discharge test

will not alter any accident analysis assumptions, initial

conditions, or results. Consequently, it does not have any effect on

the margin of safety. Therefore, this change does not involve a

significant reduction in a margin of safety.

(2) The proposed change to delete the requirement to demonstrate

that the DG can achieve and maintain steady state operation within

10 seconds is not an accident analysis assumption. The accident

analysis assumption that the DG reaches minimum conditions to accept

load within 10 seconds is preserved. Consequently, it does not have

any effect on the margin of safety. Therefore, this change does not

involve a significant reduction in a margin of safety.

(3) The editorial changes are non-technical and therefore 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.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

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

NRC Section Chief: Robert A. Gramm.

[[Page 35213]]

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

Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

Date of amendment request: May 14, 1999.

Brief description of amendments: The proposed license amendments

would change the name of the CPSES licensee from ``Texas Utilities

Electric Company'' to ``TXU Electric Company'' in the Facility

Operating Licenses of CPSES, Units 1 and 2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequence of an accident previously evaluated?

No. This request involves an administrative change only. The

Operating Licenses (OLs) are being changed to reference the new

corporate name of the licensee. No actual plant equipment or

accident analyses will be affected by the proposed change.

Therefore, TU [Texas Utilities] Electric concludes that this request

will have no impact on the possibility of any type of accident,

whether new, different or previously evaluated.

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

different kind of accident from any accident previously evaluated?

No. This request involves an administrative change only. The OLs

are being changed to reference the new corporate name of the

licensee. No actual plant equipment or accident analyses will be

affected by the proposed change and no failure modes not bounded by

previously evaluated accidents will be created. Therefore, TU

Electric concludes that this request will have no impact on the

possibility of any type of accident, whether new, different or

previously evaluated.

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

margin of safety?

No. Margin of safety is associated with confidence in the

ability of the fission product barriers (i.e., fuel and fuel

cladding, Reactor Coolant System pressure boundary, and containment

structure) to limit the level of radiation dose to the public. This

request involves an administrative change only. The OLs are being

changed to reference the new corporate name of the licensee. No

actual plant equipment or accident analyses will be affected by the

proposed change. Additionally, the proposed change will not relax

any criteria used to establish safety limits, will not relax any

safety systems settings, or will not relax the bases for any

limiting conditions of operation. Therefore, this request will not

impact margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

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

NRC Section Chief: Robert A. Gramm.

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

Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

Date of amendment request: May 24, 1999.

Brief description of amendments: The proposed license amendments

would remove several cycle-specific parameter limits from the Technical

Specifications (TSs) and add parameter limits to the Core Operating

Limits Report. In addition, the core safety limit curves would be

replaced with safety limits more directly applicable to the fuel and

fuel cladding fission product barriers. The affected TSs are: (1) TS

2.0, ``Safety Limits (SLs)''; (2) TS 3.3.1, ``Reactor Trip System

Instrumentation Setpoints''; (3) TS 3.4.1, ``RCS pressure temperature

and flow from Nucleate Boiling (DNB) Limits''; and (4) TS 5.6.5, ``Core

Operating Limits Report.''

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. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes remove cycle-specific parameter limits from

the Technical Specifications, add them to the list of limits

contained in the Core Operating Limits Report (COLR), and revise the

Administrative Controls section of the Technical Specifications. The

proposed changes also insert the original minimum RCS [reactor

coolant system] flow limits into the Technical Specifications. The

changes do not, by themselves, alter any of the parameter limits.

The changes are administrative in nature and have no adverse effect

on the probability of an accident or on the consequences of an

accident previously evaluated. The removal of parameter limits from

the Technical Specifications does not eliminate the requirement to

comply with the parameter limits.

The parameter limits in the COLR may be revised without prior

NRC approval. However, [Technical] Specification 5.6.5c continues to

ensure that the parameter limits are developed using NRC-approved

methodologies and that applicable limits of the safety analyses are

met. While future changes to the COLR parameter limits could result

in event consequences which are either slightly less or slightly

more severe than the consequences for the same event using the

present parameter limits, the differences would not be significant

and would be bounded by the requirement of specification 5.6.5c to

meet the applicable limits of the safety analysis.

Based on the above, addition of the minimum RCS flow limit into

the Technical Specifications, removal of the parameter limits from

the Technical Specifications and the addition of the described

limits in the COLR, thus allowing revision of the parameter limits

without prior NRC approval, has no significant effect on the

probability or consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated?

The proposed changes add the minimum RCS flow limit into the

Technical Specifications, remove certain parameter limits from the

Technical Specifications and add these limits to the list of limits

in the COLR, thus removing the requirement for prior NRC approval of

revisions to those parameters. The changes do not add new hardware

or change plant operations and therefore cannot initiate an event

nor cause an analyzed event to progress differently. Thus, the

possibility of a new or different kind of accident is not created.

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

margin of safety?

The margin of safety is the difference between the acceptance

criteria and the associated failure values. The proposed changes do

not affect the failure values for any parameter. Through the

accident analyses, all applicable limits (i.e., relevant event

acceptance criteria as described in the NRC-approved analysis

methodologies) are shown to be satisfied; therefore, there is no

impact on event acceptance criteria. Because neither the failure

values nor the acceptance criteria are affected, the proposed change

has no effect on the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

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

NRC Section Chief: Robert A. Gramm.

[[Page 35214]]

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: May 5, 1999.

Description of amendment request: The proposed change modifies the

Technical Specifications (TS) to enhance limiting conditions for

operation and surveillance requirements relating to the Standby Liquid

Control (SLC) system and incorporates certain provisions of NRC's rule

on anticipated transients without scram (ATWS) (10CFR50.62). The change

involves the use of enriched boron in the SLC system and improves upon

other aspects of the TS for this system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment, will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change deletes the requirement for standby liquid

control (SLC) system operability during refueling and modifies the

conditions for allowing the system to be inoperable when shutdown.

This change also permits changing the reactor mode switch to the

``Run'' or ``Startup/Hot Standby'' position to test mode switch

interlock functions while the SLC system is inoperable. To allow

testing of instrumentation associated with the reactor mode switch

interlock functions, compensatory measures are provided for assuring

that no core alterations are in progress and that all control rods

remain fully inserted in core cells containing one or more fuel

assemblies. These compensatory measures ensure that no credible

mechanisms for an inadvertent criticality are introduced by

administratively controlling the required functions of the reactor

mode switch interlocks. Control rods are not required to be inserted

in empty core cells (i.e., those containing no fuel) because, with

one or more cells in this configuration, the overall shutdown margin

is actually greater than when all control rods and all fuel

assemblies are inserted.

The SLC system is not assumed in the initiation of any

previously evaluated events and therefore the proposed change will

not significantly increase the probability or consequences of a

previously analyzed accident. The SLC system is not assumed to

operate in the mitigation of any previously analyzed accidents which

are assumed to occur during shutdown or refueling conditions. This

change will not result in operation that will significantly increase

the probability of initiating an analyzed event. This change will

not alter assumptions relative to mitigation of an accident or alter

the operation of process variables, structures, systems, or

components as described in the final safety analysis report.

VY has determined that the proposed change to increase the

standby liquid control system reactivity control capacity using a

borated water solution enriched in the boron-10 isotope effectively

increases the rate of injection of neutron absorber and does not

alter the function of the system, method of operation or dual train

configuration. The system response time to an anticipated transient

without scram (ATWS) event has been reduced as the increased boron-

10 enrichment of the solution provides faster negative reactivity

insertion, thus reducing the consequences of the ATWS event. The SLC

system is not credited in any of the design basis accident analyses

and, as such, is considered to provide only an additional mitigative

feature in the event of an accident. The SLC system sodium

pentaborate solution concentration and flow rate required by the

ATWS rule (10CFR50.62) for reactivity control independent of the

control rods are not reduced from the values previously evaluated

and presented in the Vermont Yankee Technical Specifications. The

addition of enriched boron provides a shutdown margin greater than

the previously calculated shutdown reactivity control capacity, and

the change does not affect the probability of an ATWS event.

Therefore, this change will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment, will not create the

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

previously evaluated.

The proposed change modifies the modes of applicability for the

SLC system. Included in this change is allowance to permit changing

the reactor mode switch to the ``Run'' or ``Startup/Hot Standby''

position to test mode switch interlock functions while the SLC

system is inoperable. Precautions are taken when manipulating the

mode switch to one of these positions to maintain all control rods

fully inserted in core cells containing at least one fuel assembly

and to not allow any core alterations. These two provisions

eliminate the possibility of introducing any credible mechanisms for

inadvertent criticality. The proposed change will not involve a

physical alteration of the plant (no new or different type of

equipment will be installed) or changes in methods governing normal

plant operation. The proposed change will not eliminate any valid

requirements necessary for safe operation.

VY has determined that the proposed change does not create the

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

previously evaluated because the proposed change involves a system

whose function is to provide an additional (backup) mitigative

shutdown capability and no system modifications are made.

The addition of enriched boron does not affect any system or

component that could initiate an accident. Thus, no new or different

type of accident is created.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

VY has determined that the proposed change does not involve a

significant reduction in a margin of safety. The proposed change

would remove the backup to the available reactivity control systems

when the reactor is in a shutdown or refueling condition. However,

this backup is not considered in the margin of safety when

determining the required reactivity for shutdown and refueling

events. This change will have no impact on any safety analysis

assumptions.

Included in this change is allowance to permit changing the

reactor mode switch to the ``Run'' or ``Startup/Hot Standby''

position to test mode switch interlock functions while the SLC

system is inoperable. The margin of safety will not be reduced

during such testing of interlock functions with the SLC system

inoperable because compensatory measures have been added to ensure

that no credible mechanisms for inadvertent criticality exist with

the reactor mode switch in other than the ``Shutdown'' or ``Refuel''

positions.

The use of enriched boron in the SLC system sodium pentaborate

solution actually increases the capability of the SLC system to

achieve cold shutdown; thus, no margin of safety is reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

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

NRC Section Chief: James W. Clifford.

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of amendment request: June 3, 1999.

Description of amendment request: The request is to amend the

operating license such that the name of the licensee is changed from

Washington Public Power Supply System to Energy Northwest. The name of

the facility will be changed from WPPS Nuclear Project No. 2 to WNP-2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

[[Page 35215]]

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This request involves an administrative change only. The

Operating License (OL) is being changed to reference the new name of

the licensee. No actual plant equipment or accident analyses will be

affected by the proposed change. Therefore, this request will have

no impact on the probability or consequence of any type of accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This request involves an administrative change only. The OL is

being changed to reference the new name of the licensee. No actual

plant equipment or accident analyses will be affected by the

proposed change and no failure modes not bounded by previously

evaluated accidents will be created. Therefore, this request will

have no impact on the possibility of any new type of accident: new,

different, or previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Margin of safety is associated with the confidence in the

ability of the fission product barriers (i.e., fuel and fuel

cladding, Reactor Coolant System pressure boundary, and containment

structure) to limit the level of radiation dose to the public. This

request involves an administrative change only. The OL is being

changed to reference the new name of the licensee.

No actual plant equipment or accident analyses will be affected

by the proposed change. Additionally, the proposed change will not

relax any criteria used to establish safety limits, will not relax

any safety system settings, or will not relax the bases for any

limiting conditions of operation. Therefore, this request will not

impact the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005-3502.

NRC Section Chief: Stephen Dembek.

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

Generating Station, Coffey County, Kansas

Date of amendment request: June 10, 1999.

Description of amendment request: The amendment would revise

Technical Specification Table 3.3-4, Functional Unit 7.b., Automatic

Switchover to Containment Sump (Refueling Water Storage Tank Level--

Low-Low) to reflect the results of calculations that were performed for

the associated instrumentation setpoints to consider the density

variations due to temperature and boric acid concentration.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The protection system performance will remain within the bounds

of the previously performed accident analysis. The protection

systems will continue to function in a manner consistent with the

plant design basis. The proposed changes will not affect any of the

analysis assumptions for any of the accidents previously evaluated,

since the changes are consistent with the setpoint methodology and

ensure adequate margin to the Safety Analysis Limit. The proposed

changes will not affect any event initiators nor will the proposed

changes affect the ability of any safety related equipment to

perform its intended function. There will be no degradation in the

performance of nor an increase in the number of challenges imposed

on safety related equipment assumed to function during an accident

situation. There will be no change to normal plant operating

parameters or accident mitigation capabilities.

Therefore these changes do 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.

There are no changes in the method by which any safety related

plant system performs its safety function. The normal manner of

plant operation remains unchanged, and no new equipment is being

introduced. The increase in the RWST [refueling water storage tank]

Level Low-Low Allowable Value still provides acceptable margin

between the nominal Trip Setpoint and Allowable Value while taking

into account a temperature and boric acid density correction. The

change in Allowable Value does not impact the systems capability to

perform an ECCS [emergency core cooling system] switchover from

injection to cold leg recirculation since the nominal Trip Setpoint

remains the same. The change in Allowable Value also will not affect

injection or recirculation of the Containment Spray System.

No new accident scenarios, transient precursors, failure

mechanisms, or limiting single failures are introduced as a result

of the proposed changes. Therefore, the proposed change does not

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

any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not affect the acceptance criteria for

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

Limit. There will be no effect on the manner in which safety limits

or Engineered Safety Features Actuation System settings are

determined nor will there be any affect on those plant systems

necessary to assure the accomplishment of protection functions.

Therefore, there will be no impact on any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

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

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

NRC Section Chief: Stephen Dembek.

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

Generating Station, Coffey County, Kansas

Date of amendment request: June 11, 1999.

Description of amendment request: The amendment would revise

Technical Specification 3.7.1.6, ``Steam Generator Atmospheric Relief

Valves,'' and its associated Bases to (1) require four atmospheric

relief valves (ARVs) to be operable; (2) eliminate the use of

``required'' in the action statements; (3) provide action statements to

address inoperability of two ARVs and three or more ARVs due to causes

other than excessive leakage; and (4) limit the Limiting Condition for

Operation (LCO) 3.0.4 exception to one inoperable ARV.

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.

Revising the LCO to require four ARVs to be OPERABLE rather than

three; eliminating

[[Page 35216]]

``required'' from the Actions; adding a new ACTION for three or more

ARVs inoperable; and limiting the LCO 3.0.4 exception to one ARV

inoperable constitute more restrictive changes from the current

Technical Specifications. The proposed changes do not affect

initiating mechanisms or mitigation capabilities associated with

SGTR [steam generator tube rupture] events analyzed in Chapter 15 of

the Updated Safety Analysis Report. The proposed changes impose more

stringent requirements to ensure that ARV OPERABILITY is maintained

consistent with the safety analysis and licensing basis, and also to

address all potential single failure scenarios. Therefore these

changes do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

With two ARVs inoperable, the allowed outage time for

restoration of all but one ARV to OPERABLE status is changed from 24

hours to 72 hours. The existing specification allows one valve to be

inoperable indefinitely and with one required ARV inoperable, the

allowed outage time for restoration is seven days. By modifying the

LCO to require four ARVs to be OPERABLE, an allowed outage time of

72 hours is more restrictive than the existing specification.

Therefore, revising the allowed outage time from 24 hours to 72

hours is acceptable based on a more restrictive allowed outage time

from the existing specification and the low probability of an event

requiring decay heat removal occurring during the restoration period

that would require the ARVs. With respect to Reactor Coolant System

cooldown for SGTR accident mitigation, the increase in time is

acceptable based on the low probability of a SGTR event occurring

during the restoration period and the low probability of a SGTR

event in conjunction with the failure of the turbine bypass system

(i.e., loss of offsite power). Therefore, this change in allowed

outage time does not result in a significant increase in the

probability or consequences of previously analyzed accidents.

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

or different kind of accident from any accident previously

evaluated.

There are no hardware changes nor are there any changes in the

method by which any safety related plant system performs its safety

function. Revising the LCO to require four ARVs to be OPERABLE

rather than three; eliminating ``required'' from the Actions; adding

a new ACTION for three or more ARVs inoperable; and limiting the LCO

3.0.4 exception to one ARV inoperable will not impact the normal

method of plant operation. The proposed changes ensure operation of

the plant remains consistent with analysis assumptions. No new

accident scenarios, transient precursors, failure mechanisms, or

limiting single failures are introduced as a result of the proposed

changes. Based on the above discussion, the proposed change does not

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

any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not affect the acceptance criteria for

any analyzed event. There will be no effect on the manner in which

safety limits or limiting safety system settings are determined nor

will there be any affect on those plant systems necessary to assure

the accomplishment of protection functions. The proposed changes

ensure operation of the plant consistent with the analysis

assumptions. Therefore, there will be no impact on any margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

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

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

NRC Section Chief: Stephen Dembek.

Previously Published Notice of Consideration of Issuance of

Amendment to Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie

Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: May 24, 1999.

Description of amendment request: Clarify nonconservative wording

of Technical Specification (TS) 3/4,5,1, ``Safety Injection Tanks,''

and revise TS 3/4.5.2, ``ECCS Subsystems--Tavg Greater Than or Equal to

325 degrees F,'' to align their associated surveillance requirements

with the intent and design bases requirements intended to be verified.

Date of publication of individual notice in the Federal Register:

June 10, 1999 (64 FR 31322).

Expiration date of individual notice: June 25, 1999.

Local Public Document Room location: Indian River Junior College

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

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

Date of application for amendment: April 12, 1999.

Brief description of amendment: The amendment is a temporary

amendment change effective until September 30, 1999, which revises

Technical Specification 3.7.8, ``Ultimate Heat Sink (UHS),'' to permit

an 8-hour delay in the UHS temperature restoration period prior to

entering the plant shutdown required actions.

Date of issuance: June 4, 1999.

Effective date: June 4, 1999.

Amendment No.: 183.

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

Technical Specifications.

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

24193).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 4, 1999.

[[Page 35217]]

No significant hazards consideration comments received: No.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

Commonwealth Edison Company, Docket No. 50-249, Dresden Nuclear Power

Station, Unit 3, Grundy County, Illinois

Date of application for amendment: May 5, 1999.

Brief description of amendment: The amendment removes the safety

valve function of the Target Rock safety/relief valve from Technical

Specifications (TS) Section 3.6.E and moves the reactor coolant system

safety valve lift pressure setpoints from TS Section 3.6.E to TS

Section 4.6.E.

Date of issuance: June 4, 1999.

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

within 30 days from the date of issuance.

Amendment No.: 168.

Facility Operating License No. DPR-25: The amendment revised the

Technical Specifications.

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

27824).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 4, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

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

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

Date of application for amendment: December 7, 1998, as

supplemented May 12, 1999.

Brief description of amendment: The amendment revises Technical

Specification 4.13A.2.a. to allow a one-time extension of the steam

generator (SG) inspection interval. In addition, the amendment would

remove the requirement of receiving NRC concurrence on the proposed SG

examination program in TS 4.13C.1.

Date of issuance: June 9, 1999.

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

30 days.

Amendment No.: 201.

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

Technical Specifications.

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

FR 6694).

The May 12, 1999, supplemental letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 9, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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: August 29, 1996, as supplemented January

8, 1998.

Brief description of amendment: The proposed changes revise

requirements prescribed in Technical Specification Surveillance

Requirement 3.3.1.1.8 and allow River Bend to increase the interval

between whole core traversing in-core probe to local power range

monitor calibrations from 1,000 megawatt days per ton (MWD/T) to 2,000

MWD/T.

Date of issuance: June 11, 1999.

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

30 days from the date of issuance.

Amendment No.: 107.

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55032).

The January 8, 1998, letter provided additional information that

did not change the scope of the original application and the initial

proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 11, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: April 30, 1998.

Brief description of amendment: The amendment revises the

definition of quadrant power tilt to clearly allow the use of either

the incore detectors or the excore detectors for determining quadrant

power tilt.

Date of issuance: June 10, 1999.

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

within 30 days from the date of issuance.

Amendment No.: 197.

Facility Operating License No. DPR-51: Amendment revised the

Technical Specifications.

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

FR 6694).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 10, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

No. 2, Pope County, Arkansas

Date of application for amendment: April 9, 1999.

Brief description of amendment: The proposed amendment modifies the

Technical Specifications (TSs) to add Limiting Condition for Operation

3.0.6 and its associated Bases. This change allows equipment that has

been removed from service or declared inoperable in compliance with the

TS Action statement to be returned to service under administrative

controls solely to perform testing required to demonstrate its

operability or the operability of other equipment. The proposed change

is consistent with TS 3.0.5 as discussed in NUREG-1432, Revision 1,

``Standard Technical Specifications for Combustion Engineering

Plants.'' TS 3.0.2 is also modified to reflect that TS 3.0.6 is an

exception to TS 3.0.2.

Date of issuance: June 7, 1999.

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

within 30 days from the date of issuance: June 7, 1999.

Amendment No.: 207.

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

Technical Specifications.

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

24196).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 7, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

[[Page 35218]]

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

Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana

Date of amendment request: October 1, 1998, as supplemented by

letters dated March 25 and May 6, 1999.

Brief description of amendment: The amendment modifies Technical

Specification (TS) 3.3.3.7.3 and Surveillance Requirement 4.3.3.7.3 for

the broad range gas detection system at Waterford 3. In addition, TS

Bases 3/4.3.3.7 has been changed to reflect the new system.

Date of issuance: June 3, 1999.

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

within 90 days from the date of issuance.

Amendment No.: 151.

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

Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64114).

The March 25 and May 6, 1999, letters provided clarifying

information that did not change the scope of the original application

and the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 3, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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

Station, Unit 3, St. Charles Parish, Louisiana

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

letter dated April 16, 1999.

Brief description of amendment: The amendment removes certain

administrative controls from the Waterford 3 Technical Specifications

and instead relies on the requirements of the new Entergy common

Quality Assurance Program Manual and the change controls of Title 10 of

the Code of Federal Regulations, Section 50.54(a).

Date of issuance: June 16, 1999.

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

60 days from the date of issuance.

Amendment No.: 152.

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

Technical Specifications.

Date of initial notice in Federal Register: February 26, 1999 (64

FR 9192).

The April 16, 1999, letter provided clarifying information that did

not change the scope of the original application and expand the initial

proposed no significant hazards consideration determination as

published in the Federal Register notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: March 9, 1999.

Brief description of amendment: This amendment modifies the

Technical Specifications to increase the inservice inspection interval,

and reduces the scope of volumetric and surface examinations for the

reactor coolant pump flywheels.

Date of issuance: June 8, 1999.

Effective date: June 8, 1999.

Amendment No.: 232.

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

Technical Specifications.

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

24196).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: September 30, 1998.

Brief description of amendment: The amendment corrected the

description of the reactor coolant system leakage detection capability

of the reactor building atmosphere gaseous radioactivity monitor in the

Improved Technical Specification Bases and the Final Safety Analysis

Report.

Date of issuance: June 14, 1999.

Effective date: June 14, 1999.

Amendment No.: 179.

Facility Operating License No. DPR-31: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64116).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 14, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal River, Florida 34428.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: December 3, 1996.

Brief description of amendment: The amendment incorporates certain

improvements from the Standard Technical Specifications for Babcock and

Wilcox plants (NUREG-1430).

Date of issuance: June 15, 1999.

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

60 days.

Amendment No.: 211.

Facility Operating License No. DPR-50: Amendment revised the

Technical Specifications.

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

FR 66708).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 15, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (Regional Depository) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa

Date of application for amendment: January 22, 1999.

Brief description of amendment: Revises Technical Specification

(TS) Section 4.3, ``Fuel Storage,'' by updating the criticality

requirements (k-infinity and U-235 enrichment limits) for storage of

fuel assemblies in the spent fuel racks. This change would allow for

storage of nuclear fuel assemblies with new designs, including GE-12

with a 10X10 pin array.

Date of issuance: June 8, 1999.

Effective date: June 8, 1999.

Amendment No.: 226.

Facility Operating License No. DPR-49: The amendment revised the

Technical Specifications.

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

FR 9192).

[[Page 35219]]

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa

Date of application for amendment: October 15, 1998, as

supplemented on December 21, 1998.

Brief description of amendment: Revise the Technical Specifications

(TS) by adding a new TS 3.7.9, ``Control Building/Standby Gas Treatment

System Instrument Air System,'' and revises (TS) 3.6.1.3, ``Primary

Containment Isolation Valves,'' Condition E.

Date of issuance: June 9, 1999.

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

within 30 days.

Amendment No.: 227.

Facility Operating License No. DPR-49: The amendment revised the

Technical Specifications.

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

(64FR9193).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 9, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401.

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: April 19, 1999.

Brief description of amendments: The amendments revise Technical

Specification (TS) 3/4.8.1.2, ``Electrical Power Systems, Shutdown,''

and its associated bases to provide a one-time extension of the 18-

month surveillance interval for specific surveillance requirements

associated with the emergency diesel generators for Units 1 and 2. The

surveillances will be performed prior to the first entry into Mode 4

following the current plant shutdown. In addition, for Unit 2 only, a

minor administrative change is included to delete a reference to TS

4.0.8, which is no longer applicable. For Unit 1 only, an editorial

change is made to add the word ``or'' to action statement 3.8.1.2.

Date of issuance: June 8, 1999.

Effective date: June 8, 1999, with full implementation within 45

days.

Amendment Nos.: 228 and 211.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 29, 1999 (64 FR

23129).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

Date of application for amendment: May 15, 1998, as supplemented by

letters dated September 25, October 13, December 9 (two letters), 1998;

January 11, April 1, and April 22, 1999.

Brief description of amendment: This amendment changes Technical

Specification (TS) 5.5, ``Storage of Unirradiated and Spent Fuel,'' to

reflect a planned modification to increase the storage capacity of the

spent fuel pool from 2776 to 4086 fuel assemblies. It also deletes an

inappropriate statement and reference within TS 5.5.

Date of issuance: June 17, 1999.

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

of its issuance to be implemented before spent fuel is stored within

the new high-density spent fuel rack modules authorized for

installation and use by this amendment.

Amendment No.: 167.

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

Technical Specifications.

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

FR 64973).

The September 25, October 13, December 9 (two letters) 1998,

January 11, April 1, and April 22, 1999, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

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

Nuclear Plant, Units 1 and 2, Appling County, Georgia.

Date of application for amendments: January 21, 1999, which

superseded application dated July 22, 1998.

Brief description of amendments: The amendments revise the

Technical Specifications high radiation trip setpoints for the reactor

building and the refueling floor ventilation exhaust monitors.

Date of issuance: June 9, 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--216; Unit 2--157.

Facility Operating License Nos. DPR-57 and NPF-5: Amendments

revised the Technical Specifications.

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

24200); this supersedes the original notice dated August 26, 1998 (63

FR 45529).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 9, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia.

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

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

Date of amendment request: March 30, 1999.

Brief description of amendments: The amendments deleted Technical

Specification 3/4.3.3.4, ``Meteorological Instrumentation,'' and its

associated Bases. These requirements have already been relocated to the

Technical Requirements Manual (TRM). Because the TRM is incorporated

within the South Texas Project updated final safety analysis report for

the units, changes to the relocated requirements will be controlled by

10 CFR 50.59.

Date of issuance: June 16, 1999.

Effective date: June 16, 1999, to be implemented within 30 days.

Amendment Nos.: Unit 1--111; Unit 2--98.

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

revised the Technical Specifications.

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

24201).

[[Page 35220]]

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

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

281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: February 16, 1999.

Brief Description of amendments: The amendments revise Technical

Specifications (TS) Sections 3.6, 3.9, and 3.16 and the associated

Bases for those sections for Units 1 and 2. The changes consolidate the

auxiliary feedwater cross-connect requirements by relocating the

electrical power requirements from Section 3.16 to Section 3.6. The TS

are also clarified with regard to permitting simultaneous entry into

certain conditions of operation on Units 1 and 2.

Date of issuance: June 7, 1999.

Effective date: June 7, 1999.

Amendment Nos.: 220 and 220.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments

change the Technical Specifications.

Date of initial notice in Federal Regi

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