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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued fromOctober 9, 1998, through October 23, 1998.

The last biweekly notice was published on October 21, 1998 (63 FR

56238).

[[Page 59585]]

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 December 4, 1998, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

[[Page 59586]]

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 and STN

50-529, Palo Verde Nuclear Generating Station, Units Nos. 1 and 2,

Maricopa County, Arizona

Date of application for amendment: October 6, 1998.

Description of amendment request: The proposed amendment would

clarify the power level threshold at which certain reactor protective

system (RPS) instrumentation trips must be enabled and may be bypassed,

and clarify that this level is a percentage of the neutron flux at

rated thermal power (RTP). The bypass power level, 1E-4% RTP, would be

specified as logarithmic power instead of thermal power. The intent of

(and the implementation of) the 1E-4% RTP RPS instrumentation bypass

threshold level in the technical specifications (TS) has always been

that this power level is neutron power, which would be indicated by

logarithmic power, and is not the heat transfer from the reactor core

to the coolant, including decay heat, which is the thermal power

definition in the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change would replace the words ``THERMAL POWER''

with ``logarithmic power'' for the 1E-4% rated thermal power (RTP)

level threshold in Table 3.3.1-1 footnotes (a) and (b), surveillance

requirement SR 3.3.1.7 Note 2, and Table 3.3.2-1 footnote (d) for

the reactor protective system (RPS) instrumentation. The purpose of

the 1E-4% RTP threshold is to (1) specify the power, below which,

the logarithmic power level trip is required to be operable and

surveilled, and (2) specify the power, above which, the local power

density (LPD) and departure from nucleate boiling ratio (DNBR) trips

are required to be operable. For these purposes, the appropriate

power threshold should be logarithmic power, which is the power

indicated on the logarithmic nuclear instrumentation, and not

thermal power. Thermal power is defined in TS section 1.1 as the

total reactor heat transfer rate to the reactor coolant, and would

include decay heat. Thermal power would therefore not drop to 1E-4%

RTP for a considerable period of time after shutdown, and would not

provide the plant protective function correlation required at 1E-4%

neutron RTP. However, logarithmic power, which is indicated by

neutron flux, does provide the plant protective function correlation

required at 1E-4% neutron RTP for the required reactor trips as

required by safety analyses. The logarithmic power level of 1E-4%

neutron RTP nominally correlates to the neutron flux measured by the

excore neutron instrumentation that is 1E-4% of the neutron flux at

100% RTP (3876 MWt) measured by the excore neutron instrumentation.

The proposed editorial amendment would also replace ``RTP'' with

``NRTP,'' in Table 3.3.1-1 footnotes (a) and (b), surveillance

requirement SR 3.3.1.7 Note 2, and Table 3.3.2-1 footnotes (c) and

(d). A definition would be added for NRTP (nuclear rated thermal

power) in section 1.1 as the indicated neutron flux at RTP. These

editorial clarifications will reflect the fact that the logarithmic

power level of 1E-4% is not a percentage of the ``total reactor core

heat transfer rate to the reactor coolant of 3876 MWt,'' as RTP is

defined in section TS 1.1, but is instead a percentage of the

indicated neutron flux at RTP.

An editorial change is also proposed to specify NRTP as the

``ALLOWABLE VALUE'' parameter for the high logarithmic power level

trip setpoint in Table 3.3.1-1 to correct the unintended omission of

the trip setpoint parameter during preparation of the Improved

Technical Specifications. This change will fill in the omitted

parameter with the correct parameter of NRTP that is also consistent

with the high logarithmic power trip setpoint parameter in Table

3.3.2-1.

These changes do not constitute a physical change to the Unit or

make changes in the RPS instrumentation setpoints, system logic or

manual actuation. In addition, these changes do not alter physical

plant equipment or the way in which plant equipment is operated.

This change is editorial in that it corrects the TS wording to match

the appropriate power parameter that was originally intended and

required by safety analyses, and that has been implemented since

original licensing of the PVNGS plants. 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.

The proposed change would replace the words ``THERMAL POWER''

with ``logarithmic power'' for the 1E-4% RTP level threshold in

Table 3.3.1-1 footnotes (a) and (b), surveillance requirement SR

3.3.1.7 Note 2, and Table 3.3.2-1 footnote (d) for the RPS

instrumentation. The purpose of the 1E-4% RTP threshold is to (1)

specify the power, below which, the logarithmic power level trip is

required to be operable and surveilled, and (2) specify the power,

above which, the LPD and DNBR trips are required to be operable. For

these purposes, the appropriate power threshold should be

logarithmic power, which is the power indicated on the logarithmic

nuclear instrumentation, and not thermal power. Thermal power is

defined in TS section 1.1 as the total reactor heat transfer rate to

the reactor coolant, and would include decay heat. Thermal power

would therefore not drop to 1E-4% RTP for a considerable period of

time after shutdown, and would not provide the plant protective

function correlation required at 1E-4% neutron RTP. However,

logarithmic power, which is indicated by neutron flux, does provide

the plant protective function correlation required at 1E-4% neutron

RTP for the required reactor trips as required by safety analyses.

The proposed editorial amendment would also replace ``RTP'' with

``NRTP,'' in Table 3.3.1-1 footnotes (a) and (b), surveillance

requirement SR 3.3.1.7 Note 2, and Table 3.3.2-1 footnotes (c) and

(d). A definition would be added for NRTP (nuclear rated thermal

power) in section 1.1 as the indicated neutron flux at RTP. These

editorial clarifications will reflect the fact that the logarithmic

power level of 1E-4% is not a percentage of the ``total reactor core

heat transfer rate to the reactor coolant of 3876 MWt,'' as RTP is

defined in section TS 1.1, but is instead a percentage of the

indicated neutron flux at RTP.

An editorial change is also proposed to specify NRTP as the

``ALLOWABLE VALUE'' parameter for the high logarithmic power level

trip setpoint in Table 3.3.1-1 to correct the unintended omission of

the trip setpoint parameter during preparation of the Improved

Technical Specifications. This change will fill in the omitted

parameter with the correct parameter of NRTP that is also consistent

with the high logarithmic power trip setpoint parameter in Table

3.3.2-1.

These changes do not constitute a physical change to the Unit or

make changes in the RPS instrumentation setpoints, system logic or

manual actuation. In addition, these changes do not alter physical

plant equipment or the way in which plant equipment is operated. The

proposed change does not introduce any new modes of plant operation

or new accident precursors. This change is editorial in that it

corrects the TS wording to match the appropriate power

[[Page 59587]]

parameter that was originally intended and required by safety

analyses, and that has been implemented since original licensing of

the PVNGS plants. Therefore, this change does not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change would replace the words ``THERMAL POWER''

with ``logarithmic power'' for the 1E-4% RTP level threshold in

Table 3.3.1-1 footnotes (a) and (b), surveillance requirement SR

3.3.1.7 Note 2, and Table 3.3.2-1 footnote (d) for the RPS

instrumentation. The purpose of the 1E-4% RTP threshold is to (1)

specify the power, below which, the logarithmic power level trip is

required to be operable and surveilled, and (2) specify the power,

above which, the LPD and DNBR trips are required to be operable. For

these purposes, the appropriate power threshold should be

logarithmic power, which is the power indicated on the logarithmic

nuclear instrumentation, and not thermal power. Thermal power is

defined in TS section 1.1 as the total reactor heat transfer rate to

the reactor coolant, and would include decay heat. Thermal power

would therefore not drop to 1E-4% RTP for a considerable period of

time after shutdown, and would not provide the plant protective

function correlation required at 1E-4% neutron RTP. However,

logarithmic power, which is indicated by neutron flux, does provide

the plant protective function correlation required at 1E-4% neutron

RTP for the required reactor trips as required by safety analyses.

The proposed editorial amendment would also replace ``RTP'' with

``NRTP,'' in Table 3.3.1-1 footnotes (a) and (b), surveillance

requirement SR 3.3.1.7 Note 2, and Table 3.3.2-1 footnotes (c) and

(d). A definition would be added for NRTP (nuclear rated thermal

power) in section 1.1 as the indicated neutron flux at RTP. These

editorial clarifications will reflect the fact that the logarithmic

power level of 1E-4% is not a percentage of the ``total reactor core

heat transfer rate to the reactor coolant of 3876 MWt,'' as RTP is

defined in section TS 1.1, but is instead a percentage of the

indicated neutron flux at RTP.

An editorial change is also proposed to specify NRTP as the

``ALLOWABLE VALUE'' parameter for the high logarithmic power level

trip setpoint in Table 3.3.1-1 to correct the unintended omission of

the trip setpoint parameter during preparation of the Improved

Technical Specifications. This change will fill in the omitted

parameter with the correct parameter of NRTP that is also consistent

with the high logarithmic power trip setpoint parameter in Table

3.3.2-1.

These changes do not constitute a physical change to the Unit or

make changes in the RPS instrumentation setpoints, system logic or

manual actuation. In addition, these changes do not alter physical

plant equipment or the way in which plant equipment is operated.

This change is editorial in that it corrects the TS wording to match

the appropriate power parameter that was originally intended and

required by safety analyses, and that has been implemented since

original licensing of the PVNGS plants. Therefore, this change does

not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Project Director: William H. Bateman.

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

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

Date of amendment request: October 14, 1998.

Description of amendment request: The proposed change will revise

the H. B. Robinson, Unit 2, Technical Specification (TS) on Residual

Heat Removal Isolation Valve Interlock. The requested change modifies

the acceptance criterion for surveillance requirement (SR) 3.4.14.2

from setpoint value to the analytical limit for overpressurization of

the Residual Heat Removal System.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The HBRSEP [H. B. Robinson Steam Electric Plant], Unit No. 2 TS

are proposed to be modified to increase the acceptance criterion for

Surveillance Requirement (SR) 3.4.14.2 from a RCS [reactor coolant

system] pressure of 465 psig to 474 psig. Carolina Power & Light

(CP&L) Company has evaluated the proposed Technical Specifications

(TS) change and has concluded that it does not involve a significant

hazards consideration. The conclusion is in accordance with the

criteria set forth in 10 CFR 50.92. The bases for the conclusion

that the proposed change does not involve a significant hazards

consideration is discussed below.

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change increases the acceptance criterion for the

Residual Heat Removal (RHR) System interlock from 465 psig to 474

psig. The new value of 474 psig is the analytical limit for the RHR

System interlock setpoint that corresponds to the highest RCS

pressure that is allowable in the RHR System without

overpressurizing the RHR System above its design pressure. The RHR

System interlock prohibits remote manual operation of the RHR

Pressure Isolation Valves (PIVS) from the control room when Reactor

Coolant System (RCS) pressure is greater than the RHR System

interlock setpoint to avoid inadvertent overpressurization of the

RHR System due to operator action. Operating procedures prohibit

opening of the RHR PIVs when RCS pressure is greater than 375 psig.

Therefore, the probability of overpressurization of the RHR System

resulting in a Loss-of-Coolant Accident (LOCA) is not affected by

the change. The RHR System interlock provides no actuation function

to mitigate the consequences of a LOCA as a result of open RHR PIVs

with RCS pressure greater than the RHR System interlock setpoint.

Therefore, the consequences of overpressurization of the RHR System

is not affected by the change. Therefore, the proposed change does

not involve any increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures, or components. The proposed change

increases the acceptance criterion for the RHR System interlock SR

from 465 psig to the analytical limit of 474 psig. Performance of a

SR at the new acceptance criterion does not introduce any new

accident initiation scenarios since the SR is performed at

acceptable RCS pressure conditions. Therefore, the proposed change

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

accident from any accident previously evaluated.

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

of safety?

The proposed change results in a new SR acceptance criterion

that corresponds to the analytical limit for the RHR System

interlock setpoint. The RHR System interlock is redundant to

administrative controls which prohibit opening the RHR System PIVs

under RCS pressure conditions which would overpressurize the RCS

System. Therefore, the proposed change does not result in a

significant reduction in the margin of safety.

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

this review, it appears that the three 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: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

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

Raleigh, North Carolina 27602.

[[Page 59588]]

NRC Project Director: Frederick J. Hebdon.

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

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

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units

1 and 2, Rock Island County, Illinois

Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2,

LaSalle County, Illinois

Date of application for amendment request: October 13, 1998.

Description of amendment request: The proposed amendments would

change the Dresden, Quad Cities, and LaSalle Technical Specifications

(TS) to reflect the use of Siemens Power Corporation (SPC) ATRIUM-9B

fuel. Specifically the proposed amendments incorporate the following

into the TS: (a) new methodologies that will enhance operational

flexibility and reduce the likelihood of future plant derates; (b)

administrative changes that eliminate the cycle-specific implementation

of ATRIUM-9B fuel and adopt Improved Standard Technical Specification

language where appropriate; and (c) changes to the Minimum Critical

Power Ratio (MCPR). This amendment request supplements the submittal of

August 14, 1998 (63 FR 48258). Changes in this supplement include only

a change in reference to a recently NRC-approved additive constant

uncertainty (ACU) generic methodology for ATRIUM-9B fuel (ANF-

1125(P)(A), Supplement 1, Appendix E) from Appendix D which provided an

interim value for ACU.

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.

The probability of an evaluated accident is derived from the

probabilities of the individual precursors to that accident. The

consequences of an evaluated accident are determined by the

operability of plant systems designed to mitigate those

consequences. Limits have been established consistent with NRC

approved methods to ensure that fuel performance during normal,

transient, and accident conditions is acceptable. These changes do

not affect the operability of plant systems, nor do they compromise

any fuel performance limits.

a. Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1 and 2)

The Reference 1 [ANF-91-048(P)(A), Supplement 1 and Supplement

2, ``BWR Jet Pump Model Revision for RELAX,'' October 1997 and NRC

SER, ``Review of Siemens Topical Report ANF-91-048(P), BWR Jet Pump

Revisison for RELAX (TAC No M995381), T.H. Essig to H.D. Curet,

September 19, 1997] methodology to be added to the Technical

Specifications is used as part of the LOCA [loss-of-coolant

accident] analysis and does not introduce physical changes to the

plant. The Reference 1 revised jet pump model changes the

calculational behavior of the jet pump under reversed drive flow

conditions. The revised jet pump model methodology makes the LOCA

model behave more realistically and calculates small break LOCA PCTs

[peak cladding temperature] that are comparable to the large break

LOCA results. Therefore, this change only affects the methodology

for analyzing the LOCA event and determining the protective APLHGR

[average planar linear heat generation rate] limits. The Technical

Specification requirements for monitoring APLHGR are not affected by

this change. The revised method will result in higher APLHGR limits,

thus the SPC fuel will be allowed to operate at higher nodal powers.

The approved methodology, however, still protects the fuel

performance limits specified by 10 CFR 50.46. Therefore, the

probability or consequences of an accident previously evaluated will

not change.

b. Addition of SPC Generic Methodology for Application of ANFB

[Advanced Nuclear Fuel for Boiling Water Reactors] Critical Power

Correlation to Non-SPC Fuel (Quad Cities Units 1 and 2 and LaSalle

Units 1 and 2)

The probability or consequences of a previously evaluated

accident are not increased by adding Reference 3 [EMF-1125(P)(A),

Supplement 1 Appendix C, ``ANFB Critical Power Correlation

Application for Coresident Fuel,'' August 1997, and NRC SER,

``Acceptance for Referencing of Licensing Topical Report EMF-

1125(P), Supplement 1 Appendix C, ``ANFB Critical Power Correlation

Application for Co-Resident Fuel,'' J.E. Lyons to R. A. Copeland,

May 9, 1997] to Section 6.9.A.6.b of the Quad Cities Technical

Specifications and Bases Section 2.1.2 and Section 6.6.A.6.b of the

LaSalle Technical Specifications. Reference 3 determines the

additive constants and the associated uncertainty for application of

the ANFB correlation to the coresident GE [General Electric Co.]

fuel. Therefore, it provides data that is used in the determination

of the MCPR Safety Limit. This approved methodology for applying the

ANFB critical power correlation to the GE fuel will protect the fuel

from boiling transition. Operational MCPR limits will also be

applied to ensure that the MCPR Safety Limit is protected during all

modes of operation and anticipated operational occurrences. Because

Reference 3 contains conservative methods and calculations and

because the operability of plant systems designed to mitigate any

consequences of accidents have not changed, the probability or

consequences of an accident previously evaluated will not increase.

c. Addition of SPC Topical for Revised ANFB Correlation Uncertainty

(Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units 1

and 2)

The probability or consequences of a previously evaluated

accident are not increased by adding Reference 7 [ANF-1125(P),

Supplement 1, Appendix E, ``ANFB Critical Power Correlation

Determination of ATRIUM-9B Additive Constant Uncertainties,'' and

NRC SER, ``Acceptance for Referencing of Licensing Topical Report

ANF-1125(P), Supplement 1, Appendix E, ``ANFB Critical Power

Correlation Determination of ATRIUM-9B Additive Constant

Uncertainties'' (TAC No. MA2437), T.H. Essig to H.D. Curet,

September 23, 1998] to Section'' 6.9.A.6.b of the Quad Cities and

Dresden Technical Specifications and Bases Section 2.1.2 and Section

6.6.A.6.b of the LaSalle Technical Specifications. Reference 7

documents the additive constant uncertainty for the SPC ATRIUM-9B

fuel design with an internal water channel. This methodology is used

to determine an input to the MCPR Safety Limit calculations, which

ensures that at least 99.9 percent of the fuel rods avoid transition

boiling during normal operation as well as anticipated operational

occurrences. This change does not require any physical plant

modifications, physically affect any plant components, or entail

changes in plant operation. This methodology for determining the

ATRIUM-9B additive constant uncertainty for the MCPR Safety Limit

calculation will continue to support protecting the fuel from

boiling transition. Operational MCPR limits will be applied to

ensure the MCPR Safety Limit is not violated during all modes of

operation and anticipated operational occurrences. Therefore, no

individual precursors of an accident are affected and the

operability of plant systems designed to mitigate the probability or

the consequences of an accident previously evaluated is not affected

by these changes.

d. Change to Minimum Critical Power Ratio Safety Limit (Quad Cities

Units 1 and 2, Dresden Unit 3, and LaSalle Units 1 and 2)

Changing the MCPR Safety Limit at Quad Cities Units 1 and 2,

Dresden Unit 3, and LaSalle Units 1 and 2 will not increase the

probability or the consequences of an accident previously evaluated.

The MCPR Safety Limits for Quad Cities Units 1 and 2, Dresden Unit

3, and LaSalle Units 1 and 2 are anticipated to be conservative and

acceptable for future cycles. Cycle specific MCPR Safety Limit

calculations will be performed, consistent with SPC's approved

methodology, to confirm the appropriateness of the MCPR Safety

Limit. Additionally, operational MCPR limits will be applied that

will ensure the MCPR Safety Limit is not violated during all modes

of operation and anticipated operational occurrences. The MCPR

Safety Limits are being set at the CPR [critical power ratio] value

where less than 0.1 percent of the rods in the core are expected to

experience boiling transition. These Safety Limits are expected to

be applicable for future cycles of ATRIUM-9B. Therefore the

probability or consequences of an accident will not increase.

[[Page 59589]]

e. Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel Reloads

(Quad Cities Unit 2 and Dresden Units 2 and 3)

The removal of footnotes from the Quad Cities and Dresden

Technical Specifications does not involve any significant increase

in the probability or consequences of an accident previously

evaluated. The footnotes were added to clarify that cycle specific

methods were used until the generic methodology was approved by the

NRC. Since the NRC has approved SPC's generic methodology for

application of the ANFB correlation to the coresident GE fuel

(Reference 3) and SPC has addressed the concerns regarding the

database used to calculate the ATRIUM-9B additive constant

uncertainties (Reference 7), the footnotes are no longer necessary.

The removal of the Unit 2 specific ``a'' pages, 2-1a and B2-3a, in

the Quad Cities Technical Specifications is justified by the removal

of the footnotes. Therefore, removing these footnotes and ``a''

pages does not require any physical plant modifications, nor does it

physically affect any plant components or entail changes in plant

operation. Therefore, the probability or consequences of an accident

previously evaluated are not expected to increase.

f. Revision to Thermal Limit Descriptions (Quad Cities Units 1 and 2,

Dresden Units 2 and 3, and LaSalle Units 1 and 2)

The revision to the Section 3 Technical Specification

description of the APLHGR limits has no implications on accident

analysis or plant operations. The purpose of the revision is to

allow flexibility for the MAPLHGR [maximum planar linear heat

generation rate] limits and their exposure basis to be specified in

the COLR [core operating limit report] and to establish consistency

with approved methodologies currently utilized by Siemens Power

Corporation, which calculate MAPLHGR limits based on bundle or

planar average exposures. This revision also provides for

consistency in the APLHGR limit Technical Specification wording

between the ComEd BWRs. The revision to the 3.11.D SLHGR [steady

state linear heat generation rate] Technical Specification for

Dresden also has no implications on accident analysis or plant

operations. The purpose of this revision is to allow flexibility for

the LHGR [linear heat generation rate] limits and their exposure

basis to be specified in the COLR. This revision makes the Dresden

LHGR definition consistent with NUREG 1433/1434, Revision 1 wording.

The definition of the Average Planar Exposure is deleted, because

the exposure basis of the APLHGR and LHGR is being removed.

Therefore, no plant equipment or processes are affected by this

change. Thus, there is no alteration 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:

Creation of the possibility of a new or different kind of

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

that accident. New accident precursors may be created by

modifications to the plant configuration, including changes in

allowable modes of operation. This Technical Specification submittal

does not involve any modifications to the plant configuration or

allowable modes of operation. No new precursors of an accident are

created and no new or different kinds of accidents are created.

Therefore, the proposed changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

a. Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1 and 2)

The revised jet pump model methodology will be used to analyze

the LOCA for LaSalle Units 1 and 2, and does not introduce any

physical changes to the plant or the processes used to operate the

plant. This change only affects the methods used to analyze the LOCA

event and determine the MAPLHGR limits. Therefore, the possibility

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

b. Addition of SPC Generic Methodology for Application of ANFB Critical

Power Correlation to Non-SPC Fuel (Quad Cities Units 1 and 2 and

LaSalle Units 1 and 2)

Addition of the generic methodology for the application of the

ANFB critical power correlation to GE fuel in Section 6.9.A.6.b of

the Quad Cities Technical Specifications and Bases Section 2.1.2 and

Section 6.6.A.6.b of the LaSalle Technical Specifications does not

introduce any physical changes to the plant, the processes used to

operate the plant, or allowable modes of operation. This change only

involves adding an NRC approved methodology, which is used to

determine the additive constants and additive constant uncertainty

for GE fuel, to Section 6 of the Technical Specifications.

Therefore, no new precursors of an accident are created and no new

or different kinds of accidents are created.

c. Addition of SPC Topical for Revised ANFB Correlation Uncertainty

(Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units 1

and 2)

Addition of the Reference 7 methodology to Section 6.9.A.6.b of

the Quad Cities and Dresden Technical Specifications and Bases

Section 2.1.2 and Section 6.6.A.6.b of the LaSalle Technical

Specifications will not create the possibility of a new or different

kind of accident from any accident previously evaluated. This

methodology describes the calculation of an input to the MCPR Safety

Limit--the ATRIUM-9B additive constant uncertainty. This change does

not introduce any physical changes to the plant, the processes used

to operate the plant, or allowable modes of operation. Therefore, no

new precursors of an accident are created and no new or different

kinds of accidents are created.

d. Change to Minimum Critical Power Ratio Safety Limit (Quad Cities

Units 1 and 2, Dresden Unit 3, and LaSalle Units 1 and 2)

Changing the MCPR Safety Limit will not create the possibility

of a new accident from an accident previously evaluated. This change

will not alter or add any new equipment or change modes of

operation. The MCPR Safety Limit is established to ensure that 99.9

percent of the rods avoid boiling transition.

The MCPR Safety Limit is changing for Quad Cities, Dresden Unit

3 and LaSalle due to the revised ATRIUM-9B additive constants and

the ATRIUM-9B additive constant uncertainty calculated in Reference

7. The new MCPR Safety Limit for Quad Cities Units 1 and 2, Dresden

Unit 3, and LaSalle Units 1 and 2 are greater than the current

values at Quad Cities Units 1 and 2, Dresden Unit 3, and LaSalle

Units 1 and 2 and are being increased now in anticipation of

bounding future reloads of ATRIUM-9B. This change does not introduce

any physical changes to the plant, the processes used to operate the

plant, or allowable modes of operation. Therefore, no new accidents

are created that are different from any accident previously

evaluated.

e. Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel Reloads

(Quad Cities Unit 2 and Dresden Units 2 and 3)

The removal of the footnotes from the Quad Cities and Dresden

Technical Specifications does not create a new or different kind of

accident from any accident previously evaluated. The removal of the

footnotes does not affect plant systems or operation. The footnotes

were temporarily established to implement a conservative cycle

specific MCPR Safety Limit until the SPC generic methodology was

approved. With the approval of References 3 and 7, these footnotes

are no longer applicable. Removing these footnotes does not

introduce any physical changes to the plant, the processes used to

operate the plant, or allowable modes of operation. The removal of

the Unit 2 specific ``a'' pages, 2-1a and B2-3a, in the Quad Cities

Technical Specifications, which is justified by the removal of the

footnotes, also does not create a new or different kind of accident

from any accident previously evaluated.

f. Revision to Thermal Limit Descriptions (Quad Cities Units 1 and 2,

Dresden Units 2 and 3, and LaSalle 1 and 2)

The revision of the APLHGR and LHGR limit descriptions will not

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

any accident previously evaluated. This revision will not alter any

plant systems, equipment, or physical conditions of the site. This

revision allows the flexibility of the APLHGR and the LHGR limits to

be specified in the COLR and to maintain consistency with the

calculated results of methodologies currently used to determine the

APLHGR. The definition of the Average Planar Exposure is deleted,

because it is being removed from LHGR and APLHGR Technical

Specifications. This change does not introduce any physical changes

to the plant, the processes used to operate the plant, or allowable

modes of operation. Therefore this change 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 for

the following reasons:

a. Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1 and 2)

The revised jet pump model methodology, and the MAPLHGRs,

resulting from the revised jet pump methodology, will continue

[[Page 59590]]

to ensure fuel design criteria and 10 CFR 50.46 compliance. The

results of LOCA analyses performed with this methodology must

continue to comply with the requirements of 10 CFR 50.46. Therefore,

there is no significant reduction in the margin of safety.

b. Addition of SPC Generic Methodology for Application of ANFB Critical

Power Correlation to Non-SPC Fuel (Quad Cities Units 1 and 2 and

LaSalle Units 1 and 2)

The margin of safety is not decreased by adding Reference 3 to

Section 6.9.A.6.b of the Quad Cities Technical Specifications and

Bases Section 1.2 and Section 6.6.A.6.b of the LaSalle Technical

Specifications. Siemens Power Corporation methodology for

application of the ANFB Critical Power Correlation to coresident GE

fuel is approved by the NRC and is the same methodology used in the

cycle specific topicals for coresident fuel (References 4 [EMF-96-

021(P), Revision 1, Application of the ANFB Critical Power

Correlation to Coresident GE fuel for LaSalle Unit 2 Cycle 8,''

February 1996, and NRC SER, ``Safety Evaluation for Topical Report

EMF-96-021(P), Revision 1, `Application of the ANFB Critical Power

Correlation to Coresident GE Fuel for LaSalle Unit 2 Cycle 8' (TAC

NO. M94964),'' D.M. Skay to I. Johnson, September 26, 1996] and 5

[EMF-96-051(P), ``Application of the ANFB Critical Power Correlation

to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15,'' May 1996,

and NRC SER, ``Approval of Topical Report EMF-96-051(P)--Quad

Cities, Unit 2 (TAC NO. M96213),'' R. Pulsifer to I. Johnson, May

16, 1997]). The MCPR Safety Limit will continue to ensure that

greater than 99.9 percent of the rods in the core avoid boiling

transition. Additionally, operating limits will be established to

ensure the MCPR Safety Limit is not violated during all modes of

operation.

c. Addition of SPC Topical for Revised ANFB Correlation Uncertainty

(Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units 1

and 2)

The MCPR Safety Limit provides a margin of safety by ensuring

that less than 0.1 percent of the rods are expected to be in boiling

transition if the MCPR Safety Limit is not violated. This Technical

Specification amendment request proposes to insert the topical

report that describes SPC's calculation of the ATRIUM-9B additive

constant uncertainty. The new ATRIUM-9B additive constant

uncertainty calculation is conservative and is based on a larger

database than previous calculations. Because the criteria of

ensuring that 99.9 percent of the rods are expected to avoid boiling

transition has not been changed and a conservative method is used to

calculate the ATRIUM-9B additive constant uncertainty, a decrease in

the margin to safety will not occur due to adding this methodology

to the Technical Specifications. In addition, operational limits

will be established to ensure the MCPR Safety Limit is protected for

all modes of operation. This revised methodology will ensure that

the appropriate level of fuel protection is being employed.

d. Change to Minimum Critical Power Ratio Safety Limit (Quad Cities

Units 1 and 2, Dresden Unit 3, and LaSalle Units 1 and 2)

Changing the MCPR Safety Limit for Quad Cities Units 1 and 2,

Dresden Unit 3, and LaSalle Units 1 and 2 will not involve any

reduction in margin of safety. The MCPR Safety Limit provides a

margin of safety by ensuring that less than 0.1 percent of the rods

are calculated to be in boiling transition if the MCPR Safety Limit

is not violated. The proposed Technical Specification amendment

request reflects the MCPR Safety Limit results from conservative

evaluations by SPC using the ANFB critical power correlation with

the ATRIUM-9B additive constant uncertainty calculated in Reference

7.

Because a conservative method is used to apply the ATRIUM-9B

additive constant uncertainty in the MCPR Safety Limit calculation,

a decrease in the margin to safety will not occur due to changing

the MCPR Safety Limit. The revised MCPR Safety Limit will ensure the

appropriate level of fuel protection. Additionally, operational

limits will be established based on the proposed MCPR Safety Limit

to ensure that the MCPR Safety Limit is not violated during all

modes of operation including anticipated operation occurrences. This

will ensure that the fuel design safety criterion of more than 99.9

percent of the fuel rods avoiding transition boiling during normal

operation as well as during an anticipated operational occurrence is

met.

e. Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel Reloads

(Quad Cities Unit 2 and Dresden Units 2 and 3)

The removal of the cycle specific footnotes in Quad Cities and

Dresden Technical Specifications does not impose a change in the

margin of safety. These footnotes were added due to concerns

regarding the calculation of the additive constant uncertainty for

the ATRIUM-9B fuel and the cycle specific application of the ANFB

critical power correlation to coresident GE fuel in Quad Cities Unit

2 Cycle 15. Because the generic ANFB application to coresident GE

fuel MCPR methodology (Reference 3) has received NRC approval and

the topical report describing the increased database used to

calculate the additive constant uncertainties for ATRIUM-9B

(Reference 7) has also received NRC approval and both are proposed

to be added to the Technical Specifications in this amendment

request, there is no reason for the footnotes to remain. Removal of

the Unit 2 specific ``a'' pages, 2-1a and B2-3a, in the Quad Cities

Technical Specifications is justified by the removal of the

footnotes. Therefore, the removal of the ``a'' pages, 2-1a and B2-

3a, also does not impose a change in the margin of safety.

f. Revision to Thermal Limit Descriptions (Quad Cities Units 1 and 2,

Dresden Units 2 and 3, and LaSalle Units 1 and 2)

The revision to the APLHGR and LHGR limit descriptions will not

involve a reduction in the margin of safety. The methodology used to

calculate the APLHGR must comply with the guidelines of Appendix K

of 10 CFR Part 50, and the APLHGR and LHGR will still be required to

be maintained within the limits specified in the COLR. The

surveillance requirements for these two thermal limits remain

unchanged. Thus, there will be no 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: For Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021; and for LaSalle, the Jacobs Memorial Library, 815 North

Orlando Smith Avenue, Illinois Valley Community College, Oglesby,

Illinois 61348-9692.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603. NRC Project

Director: Stuart A. Richards.

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: September 30, 1998.

Description of amendment request: The proposed amendment would

increase the maximum fuel rod internal pressure in the spent fuel pool

from 1200 pounds per square inch gauge (psig) to 1300 psig by changing

the Updated Final Analysis Report (UFSAR) reference to the computer

code used to determine the fuel rod internal pressure (TACO3 computer

code would be added) in UFSAR Chapter 15. The proposed amendment would

also provide justification for not increasing the overall effective

decontamination factor for iodine as a consequence of a fuel handling

accident. In addition, the term ``fuel assembly gap gas pressure''

would be changed to ``fuel rod internal pressure'' to correct an UFSAR

error.

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

[[Page 59591]]

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 increase in maximum rod internal pressure

in the spent fuel pool from 1200 psig to 1300 psig does not result

in a significant change in the calculated overall effective

decontamination factor for iodine (described in Attachment 1) [of

the licensee's submittal]. Therefore, the continued use of an

overall effective decontamination factor for iodine of 89 can be

justified. Therefore, there is no significant increase in the dose

consequences for a fuel handling accident at Oconee Nuclear Station.

Implementation of the BAW-10183P-A (Reference 4) methodology,

which allows fuel rod internal pressure to exceed system pressure,

also increases the fuel rod pressure at spent fuel pool conditions.

The fuel is currently licensed to rod internal pressure of system

pressure plus a proprietary amount above system pressure. This

criteria represents a separate limit from the maximum internal

pressure in the spent fuel pool criteria. Thus, an increase in the

maximum rod internal pressure in the spent fuel pool does not affect

the mechanical design limit specified in Reference 4. Therefore, an

increase in the maximum internal pressure in the spent fuel pool

does not constitute a significant increase in the probability 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 fuel handling accident is the bounding accident.

Implementation of this amendment will not impact any plant systems

that are accident initiators. No other modifications are being

proposed in the plant which 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

Implementation of this amendment would not involve a significant

reduction in a margin of safety. As discussed in Attachment 1 [of

the licensee's submittal], the overall effective decontamination

factor (DF) of 522 was determined for a rod internal pressure of

1200 psig, and a DF of 443 for a rod internal pressure of 1300 psig

based on a spent fuel pool depth of 21.34 feet. Both of these

factors are well above the DF of 89 currently used in the fuel

handling accident analyses. The margin of safety is a factor of 5.

Based upon the preceding analysis, Duke proposes that ample

margin is retained to justify the continued use of a DF of 89 at a

maximum rod internal pressure of 1300 psig. 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: J. Michael McGarry III, Winston and Strawn,

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

NRC Project Director: Herbert N. Berkow.

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

Station, Unit 2, Shippingport, Pennsylvania.

Date of amendment request: September 24, 1998.

Description of amendment request: The proposed amendment would

revise technical specification (TS) 3.1.2.8 in two places to change the

term ``contained volume'' to ``usable volume.'' This change would

eliminate the potential for a non-conservative interpretation of the

specification values for the Refueling Water Storage Tank and Boric

Acid Storage System (BAT) and would eliminate the need for plant

administrative controls, which currently interpret these volumes as

usable volumes.

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 Limiting Condition for Operation (LCO) change will

assure that the Refueling Water Storage Tank (RWST) minimum usable

volume is maintained consistent with that required by accident

analysis. The safety function of the RWST will not differ in any way

from its normal operational mode. The normal operation of plant

equipment is not a precursor to any accident. Therefore, operation

of equipment under this change will not increase 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 will not change the physical plant or the

modes of plant operation defined in the operating license. The

change does not involve the addition or modification of equipment

nor does it alter the design or operation of plant systems. The

proposed change will help to ensure that the analysis value of

minimum contained volume is available, so that the RWST can perform

its safety function.

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?

RWST: The basis for TS 3.1.2.8.b is to ensure adequate water for

the Emergency Core Cooling System to respond to a Large Break Loss

Of Coolant Accident; supply the containment with cooling spray flow;

supply the containment sump with adequate water for Recirculation

Spray pump suction head concerns; and to provide adequate boron to

shut down the core. This change will ensure that the proper tank

volume is maintained to support the Design Basis Accident (DBA)

analysis.

BAT: These tanks are credited for ensuring adequate Shutdown

Margin in the event that the unit has to initiate an emergency

shutdown. Additional requirements are derived for the postulated

Anticipated Transient Without Scram event. This change will ensure

that the proper tank volume is maintained to support the DBA

analysis.

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 1500l.

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

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

NRC Project Director: Robert A. Capra.

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

Station, Unit 2, Shippingport, Pennsylvania

Date of amendment request: October 16, 1998.

Description of amendment request: The proposed amendment would

extend on a one time only basis, the surveillance interval for

technical specifications (TSs) 4.8.1.1.1.b and 4.8.1.2 from its current

due date of January 30, 1999, to the first entry into Mode 4 following

the seventh refueling outage (2R7), but not later than May 1, 1999, by

adding a new License Condition 2.C(12). The purpose of TSs 4.8.1.1.1.b

and 4.8.1.2 is to demonstrate the ability to transfer the unit power

[[Page 59592]]

supply from the unit circuit to the system circuit.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change is temporary and allows a one time extension

of the automatic transfer function 18 month surveillance requirement

specified in Surveillance Requirement (SR) 4.8.1.1.1.b. This

surveillance requirement is also referenced in SR 4.8.1.2. The

proposed surveillance interval extension will not cause a

significant reduction in system reliability nor affect the ability

of a system to perform its design function. The proposed change does

not affect the UFSAR [Updated Final Safety Analysis Report] accident

analyses since a loss of offsite power is assumed during a design

basis accident. Therefore, this change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

Extending the surveillance interval for the performance of

specific testing will not create the possibility of any new or

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

configurations, plant equipment or analyses. The UFSAR accident

analyses assume a loss of offsite power; therefore, loss of the

automatic bus transfer feature 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?

Extending the surveillance interval for the automatic transfer

function will not impact any plant safety analyses since the UFSAR

accident analyses assume the loss of offsite power. The safety

limits assumed in the accident analyses and the design function of

the equipment required to mitigate the consequences of any

postulated accidents will not be changed since only the 18 month

surveillance test interval is being extended. Based on engineering

judgment, extending the surveillance test interval for the

performance of this specific test could slightly reduce the margin

of safety derived from the required surveillances. However, past

experience has shown that the system which automatically transfers

power from the unit to the system circuit supply is reliable. The

manual transfer requirement of SR 4.8.1.1.1.b demonstrates that the

breakers relied upon for the transfer of power are functional and

provides an opportunity to identify potential equipment degradation.

The manual transfer requirement of SR 4.8.1.1.1.b will continue to

be completed within the required surveillance interval. Therefore,

the plant will be maintained within the analyzed limits and the

proposed extension will not significantly reduce the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 1500l.

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

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

NRC Project Director: Robert A. Capra.

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

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

Date of amendment request: September 22, 1998.

Description of amendment request: The proposed amendment would

delete license conditions associated with the River Bend Station (RBS)

Transamerica Delaval, Inc. (TDI) emergency diesel generators (EDGs),

which prescribe certain inspection requirements associated with various

overload conditions experienced by the EDGs. Current license

requirements were issued following publication of NUREG-1216, which

called for extensive periodic engine tear-downs as the major part of a

maintenance and surveillance program for TDI engines. The proposed

removal of license conditions appears to be consistent with the NRC's

approval of Generic Topical Report TDI-EDG-001-A ``Basis for

Modification to Inspection Requirements for Transamerica Delaval, Inc.,

Emergency Diesel Generators''. EOI currently inspects and maintains its

EDGs in accordance with Technical Requirements Manual (TRM)

surveillance requirement TSR 3.8.1.21. Periodicity of planned

inspections and maintenance are based upon the manufacturer's

recommendations for standby service.

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 the

consequences of an accident previously evaluated:

Diesel generators are not accident initiating equipment.

Elimination of the non-routine tear-downs and inspections will not

adversely affect the probability of an accident occurring. Regular

maintenance programs (which may include periodic tear-downs and

inspections) in lieu of this specific license condition would

decrease the consequences of an accident because of the availability

of the engines will increase as a result of the less frequent tear-

downs. (See Generic Topical Report TDI-EDG-001-A, ``Basis for

Modification to Inspection Requirements for Transamerica Delaval,

Inc., Emergency Diesel Generators'') Additionally, the high average

reliability of the TDI engines will not be negatively affected due

to this change. NRC research has shown there is a period of

decreased reliability immediately following intrusive tear-downs

(break-in period), followed by a long period of high reliability.

Continued monitoring and maintenance as implemented by Technical

Requirements Manual (TRM) surveillances will contribute to continued

high reliability of the EDGs.

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

from any previously evaluated:

The proposed amendment does not affect the design or function of

any plant structure, system, or component, nor does it change the

way plant systems are operated. The proposed amendment will not

cause any physical change to the plant or the design or operation of

the diesel units. This change will only affect the frequency of

tear-down inspections of the EDGs, and not the physical activities

performed during such inspections. Therefore, the removal of the

existing condition from the operating license will not create the

possibility of a new or different kind of accident from any

previously evaluated.

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

The proposed amendment does not affect parameters which would

result in a significant reduction in margin of safety. Operating

experience and data have shown increased reliability can be achieved

by eliminating unnecessary tear-down inspections, such as those

prescribed by this license condition. Maintenance of the EDGs is

presently scheduled in accordance with the vendor's recommendations.

The RBS corrective action program provides a means to evaluate

future operational events and take the appropriate actions.

Therefore, the proposed amendment does not involve a significant

decrease 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: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

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

1400 L Street, N.W., Washington, D.C. 20005.

NRC Project Director: John N. Hannon.

[[Page 59593]]

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: May 28, 1998.

Description of amendment request: This amendment requests changes

to Technical Specification 3.7.1.2 and Surveillance Requirement 4.7.1.2

for the Emergency Feedwater System. The amendment will expand and

clarify the current specification. A change to Technical Specification

Bases 3/4.7.1.2 has been included to support the 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. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

The proposed changes included in this amendment request are

being made to the Emergency Feedwater (EFW) System Technical

Specification. These changes include clarification of the LCO

[limiting conditions for operation], a 7 day allowed outage time for

an inoperable steam supply, additional ACTION requirements for

inoperable flow path(s), a requirement to test the pumps pursuant to

Specification 4.0.5, and rewording of numerous Surveillance

Requirements consistent with NUREG-1432, ``Standard Technical

Specifications Combustion Engineering Plants.''

The administrative and more restrictive changes will not affect

the assumptions, design parameters, or results of any accident

previously evaluated. The accident mitigation features of the plant

are not affected by these proposed changes. The proposed changes do

not add or modify any existing equipment. The administrative change

to test EFW pumps pursuant to the Inservice Test Program will ensure

the EFW pumps are tested against the more restrictive of the data

points required by either the safety analysis or the Inservice Test

Program. Therefore, the proposed administrative changes do not

involve a significant increase in the probability or consequences of

any accident previously evaluated.

The less restrictive changes (allowing 7 days for an inoperable

pump due to an inoperable steam supply, performing Surveillance

Requirements during other than shut down conditions, allowing the

use of actual actuation signals in addition to test signals, and

delaying the requirement to complete Surveillance Requirement ``d''

to just prior to Mode 2) will not affect the assumptions, design

parameters, or results of any accident previously evaluated. The

accident mitigation features of the plant are not affected by these

proposed changes. The proposed changes do not add or modify any

existing equipment. Therefore, the proposed less restrictive changes

do not involve a significant increase in the probability or

consequences of any accident previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

The proposed changes included in this amendment request are

being made to the EFW System Technical Specification. These changes

include clarification of the LCO, a 7 day allowed outage time for an

inoperable steam supply, additional ACTION requirements for

inoperable flow path(s), a requirement to test the pumps pursuant to

Specification 4.0.5, and rewording of numerous Surveillance

Requirements consistent with NUREG-1432. These changes do not alter

the design nor configuration of the plant. There has been no

physical change to plant systems, structures, or components. The

proposed changes will not reduce the ability of any of the safety-

related equipment required to mitigate Anticipated Operational

Occurrences or accidents. Therefore, the proposed changes will not

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

any accident previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

The proposed changes included in this amendment request are

being made to the EFW System Technical Specification. These changes

include clarification of the LCO, a 7 day allowed outage time for an

inoperable steam supply, additional ACTION requirements for

inoperable flow path(s), a requirement to test the pumps pursuant to

Specification 4.0.5, and rewording of numerous Surveillance

Requirements consistent with NUREG-1432.

The proposed change to the LCO requiring three pumps and two

flow paths be OPERABLE maintains the functionality of the EFW such

that it is capable of performing its design function as assumed in

the Updated Final Safety Analysis Report. If the functionality of

the system is not maintained, Technical Specifications require

ACTIONs be taken, within specified time limitations, to restore EFW

to OPERABLE status or shut down the reactor. This action is

consistent with the existing Technical Specification and NUREG-1432.

The allowed outage time for one inoperable steam supply has been

increased from 72 hours to 7 days in accordance with NUREG-1432.

This is acceptable due to the redundant OPERABLE steam supply, the

availability of redundant OPERABLE motor-driven EFW pumps, and the

low probability of an event requiring the inoperable steam supply.

This change is consistent (other than format) with NUREG-1432 and

has therefore been previously approved by the NRC.

The ACTION for one flow path inoperable (but capable of

delivering 100% flow) as proposed will allow a 72 hour completion

time for an inoperable flow path. This change is acceptable based on

the availability of at least two OPERABLE EFW pumps, a redundant

OPERABLE flow path capable of feeding the other steam generator and

the capability of the inoperable flow path to deliver 100% of the

required EFW flow to the affected steam generator.

The ACTION for one flow path inoperable (not capable of

delivering 100% flow) as proposed requires a unit shutdown be

initiated immediately. This change is appropriate due to the

seriousness of the condition and is acceptable due to the

availability of the remaining operable flow path to support the unit

shut down.

The ACTION for two flow paths not capable of delivering 100%

flow is the same as that for three pumps inoperable. With two flow

paths inoperable such that neither flow path is capable of

delivering 100% flow the unit is in a seriously degraded condition

just as it is with all three pumps inoperable. The ACTION as

proposed requires that immediate action be taken to restore one flow

path to OPERABLE status. This change is consistent with the intent

of the current EFW Technical Specification.

Testing pursuant to Specification 4.0.5 (Inservice Testing

Program) as proposed for Surveillance Requirement `b' will ensure

the EFW pumps are tested against the more restrictive of the data

points required by either the safety analysis or ASME Section XI.

The remaining changes to the EFW Technical Specification are

consistent (other than format) with NUREG-1432 and have therefore

been previously approved by the NRC.

Therefore, based on the above discussion, the proposed change

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: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

Street N.W., Washington, D.C. 20005-3502.

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: September 28, 1998.

Description of amendment request: In 1997 Northeast Nuclear Energy

Company (the licensee) changed the Final Safety Analysis Report (FSAR)

Section 8.7.3.1 electrical separation requirements from 12 inches to 6

[[Page 59594]]

inches. At that time, the licensee concluded that the FSAR changes did

not involve an unreviewed safety question. Therefore, the licensee did

not request a license amendment to implement the FSAR change. The

licensee has since determined that, although the changes were safe, an

unreviewed safety question was involved. Therefore, the licensee is now

requesting NRC's review and approval, through an amendment to Operating

License No. DPR-65 pursuant to 10 CFR 50.90, regarding the separation

requirement of 6 inches in Millstone Unit No. 2 FSAR (which is applied

to redundant vital cables, internal wiring of redundant vital circuits,

and associated devices).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with 10CFR50.92, NNECO [Northeast Nuclear Energy

Company] has reviewed the proposed changes and has concluded that

they do not involve a Significant Hazards Consideration (SHC). The

basis for this conclusion is that the three criteria of

10CFR50.92(c) are not compromised. The proposed changes do not

involve an SHC because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The FSAR changes reduce the minimum allowable separation between

redundant vital wires/devices of different channels from twelve

inches to six inches. Reducing the physical separation between

wires/devices does not in itself increase the probability of any

credible event that would challenge circuit operability since the

wire/device characteristics have not changed and there is no change

in the circuit the wires/devices are in. The probability that an

accident could occur due to the change in separation is not

increased since the remaining separation will still prevent adverse

channel interactions (i.e. short circuit, etc.). The six inch

standard is acceptable in accordance with IEEE standard 384-1981

[IEEE standard 384-1981, ``Standard Criteria for Independence of

Class 1E Equipment and Circuits''], sections 6.6.2 and 6.6.5, and

IEEE standard 420-1982, [IEEE standard 420-1982, ``Design Standards

and Qualification of class 1E Control Boards, panels, and Racks Used

in Nuclear Power Generating Stations''], sections 4.3.1, 4.3.2, and

4.3.3 which have been endorsed by the NRC in Regulatory Guide 1.75

[Regulatory Guide 1.75, ``Physical Independence of Electrical

Systems'']. Therefore, these changes will not significantly increase

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.

The FSAR changes reduce the minimum allowable separation between

redundant vital wires/devices of different channels from twelve

inches to six inches. The new minimum allowable separation will not

introduce any new or unanalyzed failure modes of equipment or

systems, and does not change the configuration of the plant. These

changes will not require any new or unusual operator actions, alter

the way any structure, system, or component functions and do not

alter the manner in which the plant is operated. Therefore, there

are no new or different types of failures of systems or equipment

important to safety which could cause a new or different type of

accident from any accident previously evaluated.

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

The FSAR changes reduce the minimum allowable separation between

redundant vital wires/devices of different channels from twelve

inches to six inches. The probability that a single wire/device

failure could cause the failure of redundant vital channels may be

increased. However, the new minimum allowed separation has been

found acceptable by IEEE standard 384-1981, sections 6.6.2 and

6.6.5, and IEEE standard 420-1982, sections 4.3.1, 4.3.2, and 4.3.3

which have been endorsed by the NRC in Regulatory Guide 1.75. The

new minimum allowed separation does not change any plant equipment

configuration, does not change the functionality of any equipment,

and does not change any operating setpoints. This change does not

alter the acceptance limits of the safety parameters of the accident

analyses stated in the FSAR. No new analysis assumptions are

required based on this change (e.g. common-cause failures).

Therefore, there is no impact 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: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Project Director: William M. Dean.

Public Service Electric & Gas Company, Docket No. 50-311, Salem Nuclear

Generating Station, Unit No. 2, Salem County, New Jersey

Date of amendment request: October 12, 1998.

Description of amendment request: The proposed amendment would

allow a one-time extension of the Technical Specification (TS)

surveillance interval to the end of fuel cycle 10 for certain TS

surveillance requirements (SRs). Specifically, SR 4.3.2.1.3 requires

the instrumentation response time testing of each engineered safety

features actuation system function at least once per 18 months and SRs

4.8.2.3.2.f and 4.8.2.5.2.d require that the 125 volt DC and the 28

volt DC distribution system batteries, respectively, be capacity

service tested at least once per 18 months, during shutdown.

Additionally, SR 4.8.2.5.2.c.2 requires that the 125 volt DC battery

connections be verified clean, tight, and coated with anti-corrosion

material at least once per 18 months. Because of the length of the last

outage and delays in restart, the SRs will be overdue prior to reaching

the next refueling outage (2R10). The SRs are to be completed during

the 2R10 outage, prior to returning the unit to Mode 4 (hot shutdown)

upon outage completion.

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:

4.3.2.1.3 (Instrumentation, Engineered Safety Feature Actuation

System Instrumentation)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The deferral of the surveillance requirement does not involve

any physical changes to the plant nor does it change the way the

plant is operated. Thus, the proposal does not increase the

probability of an accident previously evaluated.

The SEC [safeguard equipment control] automatic self-test

feature, the monthly functional surveillance testing and the

positive surveillance testing history provide sufficient assurance

of the operability of the system. These features also provide

assurance that a degraded condition, if it did occur, would be

detected.

Thus, it is reasonable to conclude that this proposal represents

no significant increase in the consequences of an accident

previously analyzed.

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

or different kind of accident from any accident previously

evaluated.

Deferral of the surveillance requirement does not involve any

physical changes to the plant nor does it change the way the plant

is operated.

Thus, it can be concluded that deferring the surveillance

requirement to the refueling outage cannot create the possibility of

a different kind of accident from any accident previously evaluated.

[[Page 59595]]

3. The proposed change does not involve a significant reduction

in a margin of safety.

Deferral of the surveillance requirement does not involve any

physical changes to the plant nor does it change the way the plant

is operated. The self-test feature and the monthly functional

testing will provide reasonable assurance that the SECs will remain

operable during the few weeks of deferral to the refueling outage.

Also the ability to detect a degraded condition in the SEC will not

be affected during the deferral period.

Therefore, the plant's response to accident conditions during

the period of deferral will not be affected.

Thus, it can be reasonably concluded that this proposal to amend

the Salem Unit 2 Technical Specifications, on a one-time basis, to

defer surveillance requirement 4.3.2.1.3 does not involve a

significant reduction in any margin of safety.

4.8.2.3.2.f, (Electrical Power Systems, 125 Volt D.C.

Distribution), and 4.8.2.5.2.c.2 and 4.8.2.5.2.d (Electrical Power

Systems, 28 Volt D.C. Distribution)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The deferral of the battery service tests to the refueling

outage does not involve any physical changes to the power plant or

to the manner in which the power plant is operated. Therefore, the

probability of an accident previously evaluated is not increased.

Weekly and quarterly testing and performance monitoring by the

system manager along with the current condition of the batteries

(past test results demonstrating above 100% capacity) provide

assurance that battery condition and performance will not

deteriorate during the deferral period. Other positive industry

experience for similar batteries on 24 month cycles also support

this assurance. Therefore, the consequences of a loss of power

accident will not be increased due to the deferral of the

surveillance requirements.

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

or different kind of accident from any accident previously

evaluated.

The deferral of the battery service tests to the refueling

outage does not involve any physical changes to the power plant or

to the manner in which the power plant is operated. No new failure

mechanisms will be introduced by the surveillance deferral.

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 change does not involve a significant reduction

in a margin of safety.

The deferral of the battery service tests to the refueling

outage does not involve any physical changes to the power plant or

to the manner in which the power plant is operated. Continuing

weekly and quarterly testing and performance monitoring along with

the current condition of the batteries provides assurance that

battery condition and performance will not deteriorate to an

unacceptable level during the deferral period and that any

degradation that may occur will be detected. Therefore, the plant's

response to accident conditions during the period of deferral will

not be affected.

Thus, it can be reasonably concluded that this proposal to amend

the Salem Unit 2 Technical Specifications, on a one-time basis, to

defer surveillance requirements 4.8.2.3.2.f and 4.8.2.5.2.d does not

involve a significant reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079.

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

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

NRC Project Director: Robert A. Capra.

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

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

Diego County, California

Date of amendment request: May 7, 1998.

Description of amendment request: This change would revise the

reference for obtaining the thyroid dose conversion factors used in the

definition of Dose Equivalent Iodine 131 (I-131) in Technical

Specification (TS) Section 1.1, ``Definitions'' for each plant.

Specifically, the reference to ``Table E-7 of Regulatory Guide 1.109,

Rev. 1, NRC 1977'' is to be replaced with a reference to the

International Commission on Radiological Protection Publication 30

(ICRP-30), Supplement to Part 1, Pages 192-212, Tables titled,

``Committed Dose Equivalent in Target Organs or Tissues per Intake of

Unit Activity.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change, which utilizes International Committee on

Radiological Protection (ICRP)-30 methodology for determining dose

equivalent Iodine-131, and therefore for evaluating thyroid dose

consequences, does not involve any change to the method of operation

of any plant equipment, nor does it modify any plant equipment. In

addition, utilization of the ICRP-30 Dose Conversion Factors (DCFs)

will effectively reduce calculated thyroid dose consequences of

design basis accidents, thereby decreasing the calculated thyroid

dose consequences of previously evaluated accidents.

Therefore, the proposed changes will not increase the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change does not modify the configuration of the

units, involve any change to plant equipment or change the method of

plant operation. The utilization of the ICRP methodology for

determining DCFs uses more recent data which only affects

calculations for determining thyroid dose consequences.

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

or different kind of accident from any previously evaluated

accident.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The change to utilize the ICRP methodology for determining DCFs

allows the use of more recent data which only affects calculations

for determining thyroid dose consequences. ICRP-30 is recognized in

Revision 1 of NUREG-1432, ``Standard Technical Specifications,

Combustion Engineering Plants,'' as an acceptable source document

for DCFs. The new methodology will result in more accurate DCFs that

will be used in the determination of dose consequences. Utilization

of the ICRP-30 DCFs will effectively reduce calculated thyroid dose

consequences of design basis accidents, thereby providing additional

design margin.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 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, P.O. Box 800, Rosemead, California 91770.

NRC Project Director: William H. Bateman.

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

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

Date of amendment request: September 30, 1998.

Description of amendment request: Revises Units 1 and 2 Technical

[[Page 59596]]

Specification (TS) Section 3/4.4.5, ``Steam Generator'' Surveillance

Requirements. The installation of the new Delta 94 steam generators at

the South Texas Project Units 1 and 2 necessitates changes to the steam

generator tube sample selection and inspection requirements; inservice

inspection frequencies; acceptance criteria; and inspection reporting

requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Eliminating provisions in the Technical Specifications for

applications of the voltage-based repair criteria, the F* alternate

repair criteria, and laser-welded sleeves for the Delta 94 steam

generators is an administrative adjustment, since the voltage-based

repair criteria, the F* alternate repair criteria, and laser-welded

sleeves are not applicable to the Delta 94 steam generators.

The Delta 94 steam generator tubing is designed and evaluated

consistent with the margins of safety specified in ASME Code Section

III.

The program for periodic inservice inspection of steam

generators monitors the integrity of the steam generator tubing to

ensure that there is sufficient time to take proper and timely

corrective action if tube degradation is present.

The ASME Section XI basis for the 40% through-wall plugging

limit is applicable to the Delta 94 steam generators just as it was

applicable to the Model E steam generators prior to the

implementation of voltage-based repair criteria, F* alternate repair

criteria, and laser-welded sleeves. In addition, analysis per

Regulatory Guide 1.121 (WCAP-15095/WCAP-15096) has confirmed the

applicability of the 40% plugging limit for the Delta 94 steam

generators.

The changes also clarify that inservice inspection is required

following steam generator replacement, and that inservice inspection

is not required during the steam generator replacement outage. This

is an administrative change in that it only provides clarification

of requirements written without steam generator replacement

considerations, and therefore, reduces the possibility for confusion

in the application of the subject technical specification

provisions. Therefore, these proposed 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.

Eliminating provisions in the Technical Specifications for

application of the voltage-based repair criteria, the F* alternate

repair criteria, and laser-welded sleeves to the Delta 94 steam

generators is an administrative adjustment, since the voltage-based

repair criteria, the F* alternate repair criteria, and laser-welded

sleeves are not applicable to the Delta 94 steam generators.

The changes also clarify that inservice inspection is required

following steam generator replacement, and that inservice inspection

is not required during the steam generator replacement outage. These

are administrative changes in that they only provide clarification

of requirements written without steam generator replacement

considerations, and therefore, reduce the possibility for confusion

in the application of the subject technical specification

provisions. Therefore, these proposed changes do 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 the margin of safety.

Eliminating provisions in the Technical Specifications for

applications of the voltage-based repair criteria, the F* alternate

repair criteria, and laser-welded sleeves for the Delta 94 steam

generators is an administrative adjustment, since the voltage-based

repair criteria, the F* alternate repair criteria, and laser-welded

sleeves are not applicable to the Delta 94 steam generators.

The Delta 94 steam generator tubing is designed and evaluated

consistent with the margins of safety specified in ASME Code Section

III. The program for periodic inservice inspection of steam

generators monitors the integrity of the steam generator tubing to

ensure that there is sufficient time to take proper and timely

corrective action if tube degradation is present.

The ASME Section XI basis for the 40% through-wall plugging

limit is applicable to the Delta 94 steam generators just as it was

applicable to the Model E steam generators prior to the

implementation of voltage-based repair criteria, F* alternate repair

criteria, and laser-welded sleeves. In addition, analysis per

Regulatory Guide 1.121 (WCAP-15095/WCAP-15096) has confirmed the

applicability of the 40% plugging limit for the Delta 94 steam

generators.

The changes also clarify that inservice inspection is required

following steam generator replacement, and that inservice inspection

is not required during the steam generator replacement outage. These

are administrative changes in that they only provide clarification

of requirements written without steam generator replacement

considerations, and therefore, reduce the possibility for confusion

in the application of the subject technical specification

provisions. Therefore, these proposed changes do not involve a

significant reduction in the margin of safety.

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

this review, it appears that the 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, TX 77488.

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

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.

NRC Project Director: John N. Hannon.

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

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

Date of application for amendments: September 20, 1996 (TS 96-09).

Brief description of amendments: The amendments would change the

Sequoyah Nuclear Plant (SQN) Technical Specifications by clarifying the

types of work shifts that are acceptable when considering the

requirements to ensure heavy use of overtime is not used routinely by

unit staff. The current ``8-hour day'' criteria in Section 6.2.2.g will

be expanded to include 10-hour and 12-hour allowances. In addition, the

``40-hour week'' criteria will be changed to a ``nominal 40-hour week''

to provide the necessary flexibility associated with the use of the

proposed shift durations.

Basis for proposed no significant hazards consideration

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

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

of no significant hazards consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

This change affects the requirements that ensure unit staff

personnel do not routinely incur heavy use of overtime. These

requirements are not changed by the proposed revision, but are

clarified to accommodate the various shift durations used at SQN.

The overtime usage by unit staff is not considered to be the

initiator for any postulated accident; therefore, the clarification

of associated requirements will not increase the probability of an

accident. Limiting the use of overtime by staff personnel enhances

the operation and maintenance of critical plant equipment that are

necessary to mitigate accidents. The proposed revision clarifies

these provisions, but does not reduce their adequacy. Therefore, the

proposed revision will not increase the consequences of an accident

previously evaluated.

[[Page 59597]]

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

from any previously analyzed.

This change only affects the clarification of shift durations

use by unit staff and is not associated with the initiators of

accidents. Therefore, the possibility of a new or different kind of

accident from any previously analyzed is not created by the proposed

clarifications.

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

The proposed changes do not affect plant equipment setpoints or

operating policies at SQN. The overtime provisions that ensure the

unit staff are capable to operate and maintain the plant in an

acceptable manner to provide safe operation and mitigation of

accidents is maintained by this change. Therefore, the margin of

safety is not reduced by the proposed changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon.

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

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: October 2, 1998.

Brief description of amendments: The proposed change would revise

Technical Specification (TS) 4.0.6, ``Steam Generator Surveillance

Requirements,'' to add definitions required for the F* alternate steam

generator tube plugging criterion and identify the portion of the tube

subject to the criteria.

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 supporting technical evaluation of the subject criterion

[Westinghouse WCAP-15004, listed as Reference 1 (Proprietary)],

demonstrates that the presence of the tubesheet enhances the tube

integrity in the region of the hardroll by precluding tube

deformation beyond its initial expanded outside diameter. The result

of hardrolling of the tube into the tubesheet is an interference fit

between the tube and the tubesheet. A tube rupture cannot occur

because the contact between the tube and tubesheet does not permit

sufficient movement of tube material. In a similar manner, the

tubesheet does not permit sufficient movement of tube material to

permit buckling collapse of the tube during postulated LOCA

loadings. Analysis and testing have been done to determine the

resistive strength of roll expanded tubes within the tubesheet. This

evaluation provides the basis for the acceptance criterion for tube

degradation subject to the F* criterion. The F* distance of roll

expansion is sufficient to preclude tube axial translation or

pullout from tube degradation located below the F* distance,

regardless of the extent of the tube degradation. The necessary

engagement length applicable to the Comanche Peak Unit 1 steam

generators is determined to be 1.13 inches, plus an allowance for

eddy current measurement uncertainty, based on preload analyses.

Verification that this value is significantly conservative was

demonstrated by both pullout and hydraulic proof testing.

Application of the F* criterion provides a level of protection for

tube degradation in the tubesheet region commensurate with that

afforded by RG 1.121. Leakage testing of roll expanded tubes

indicates that for roll lengths approximately equal to the F*

distance, any postulated faulted condition primary to secondary

leakage from F* tubes would be insignificant. No leakage occurred

from any of the hydraulic proof test specimens for pressures up to

and exceeding faulted condition events. The existing Technical

Specification leakage rate requirements and accident analysis

assumptions remain unchanged.

Based on the above, it is concluded that the proposed F*

criterion does not adversely impact any other previously evaluated

design basis accidents and operation of Comanche Peak Unit 1 in

accordance with the proposed license amendment does 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?

Implementation of the proposed F* criterion does not introduce

any significant changes to the plant design basis. Use of the F*

criterion does not provide a mechanism to result in an accident

initiated outside of the region of the tubesheet expansion. Even if

it is postulated that a circumferential separation of a F* tube were

to occur below the F* distance, tube structural and leakage

integrity will be maintained consistent with the assumptions of the

design basis accidents during all plant conditions. Verification of

the F* distance of non-degraded tube roll expansion prevents a

postulated separated tube from lifting out of the tubesheet during

all plant conditions. The F* criterion does not create a possibility

for simultaneous failures of multiple tubes. Any other hypothetical

accident as a result of any degradation in the expanded portion of

the tube would be bounded by the existing steam generator tube

rupture accident analysis.

Therefore, it is concluded that the proposed license amendment

does 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?

The use of the F* criterion has been demonstrated to maintain

the integrity of the tube bundle commensurate with the requirements

of RG 1.121 (intended for indications in the free span of tubes) and

the primary to secondary pressure boundary under normal and

postulated accident conditions. Acceptable tube degradation for the

F* criterion is any degradation indication in the tubesheet region,

more than the F* distance below the bottom of the transition between

the roll expansion and the unexpanded tube or the bottom of the

tubesheet (whichever is lower). The safety factors used in the

verification of the strength of the degraded tube are consistent

with the safety factors in the ASME Boiler and Pressure Vessel Code

used in steam generator design. The F* distance has been verified by

pullout and hydraulic proof testing of tubes in tubesheet simulating

collars to be greater than the length of roll expansion required to

preclude both tube pullout and significant leakage during normal and

postulated accident conditions. Resistance to tube pullout is based

upon the primary to secondary pressure differential as it acts on

the surface area of the tube, which includes the tube wall cross-

section, in addition to the inner diameter based area of the tube.

The leak testing acceptance criteria are based on the primary to

secondary leakage limit in the Technical Specifications and the

leakage assumptions used in the FSAR accident analyses.

Implementation of the proposed F* criterion will decrease the

number of tubes which must be taken out of service with tube plugs.

Plugged tubes reduce the RCS flow margin, thus implementation of the

F* alternate plugging criterion will maintain the margin of flow

that would otherwise be reduced in the event of increased plugging.

Therefore, it is concluded that the proposed change does not

result in a significant reduction in margin to plant safety as

defined in the Final Safety Analysis Report or the bases of the

Technical Specifications.

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, TX 76019.

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

Bockius, 1800 M Street, N.W., Washington, DC 20036.

[[Page 59598]]

NRC Project Director: John N. Hannon.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: September 12, 1996, as supplemented

April 24, 1997, and September 24, 1998.

Description of amendment request: The staff had previously

published a Notice of Consideration of Amendments and Proposed No

Significant Hazards Consideration Determination for the licensee's

September 12, 1996, application in the Federal Register on April 23,

1997 (62 FR 19835). As a result of the staff's requests for additional

information, the licensee supplemented its original proposal to

relocate the fire protection requirements from the Technical

Specifications (TS) to the Updated Final Safety Analysis Report (UFSAR)

by letters dated April 24, 1997, and September 24, 1998. The April 24,

1997, letter corrected two minor administrative oversights and does not

affect the No Significant Hazards Consideration Determination (NSHCD).

However, the September 24, 1998, letter revised the original

application to require the Station Nuclear Safety and Operating

Committee to submit recommended changes to the offsite review group. In

addition, a requirement was added for the establishment,

implementation, and maintenance of the Fire Protection Program and

implementing procedures. The NSHCD for these changes, as provided in

the September 24, 1998, letter, is addressed below.

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:

Since these two changes only deal with administrative

requirements, neither of these two specific changes would result in

a significant hazards consideration. Therefore, the operation of

Surry Power Station with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The probability of an accident is not increased as a result of

this Technical Specifications change request. This is an

administrative change and merely incorporates two additional

requirements for ensuring that the Fire Protection Program and

implementing procedures are appropriately established, implemented

and maintained, and that changes to the Program and implementing

procedures receive the appropriate offsite review. The consequences

of an accident previously evaluated are not increased since the

station will not be operated differently, and no physical

modifications are being made to plant systems or components.

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

from any accident previously evaluated.

A new or different type of accident is not being created since

this TS change request is administrative. As noted above, the

station will not be operated differently, and no physical

modifications are being made to plant systems or components.

Administrative revisions regarding the establishment, implementation

and maintenance of a TS requirement for a Fire Protection Program

and implementing procedures and the imposition of an offsite review

for changes thereto [do] not create a new or different type of

accident from any accident previously evaluated.

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

The margin of safety as defined in the Technical Specifications

is not reduced since system/component performance as assumed in the

existing safety analyses is not being affected by the proposed TS

change. The TS change is administrative in nature and, as such, has

no effect on station operation. The Fire Protection Program is being

retained and maintained in the UFSAR and station procedures.

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: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

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

Virginia.

NRC Project Director: Herbert N. Berkow.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed no Significant Hazards

Consideration Determination and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

Illinois Power Company, Docket, No. 50-461, Clinton Power Station,

DeWitt County, Illinois

Date of application for amendment: October 5, 1998.

Brief description of amendment request: The proposed amendment

requests deferral of the next scheduled local leak rate test for valve

1MC-042 until the seventh refueling outage.

Date of publication of individual notice in Federal Register:

October 23, 1998 (63 FR 56949).

Expiration date of individual notice: November 23, 1998.

Local Public Document Room location: Vespasian Warner Public

Library, 310 N. Quincy Street, Clinton, IL 61727.

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

Power Station, Unit 3, New London County, Connecticut

Date of amendment request: August 6, 1998, as supplemented by

letters dated September 3 and 21, 1998.

Description of amendment request: The proposed amendment allows a

one-time extension to the steam generator tube inspection surveillance

interval until the next refueling outage or July 1, 1999, whichever

date is earlier.

Date of publication of individual notice in Federal Register:

August 17, 1998 (63 FR 43964).

Expiration date of individual notice: September 16, 1998.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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.

[[Page 59599]]

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, et al., Docket No. 50-400, Shearon

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

Carolina

Date of application for amendment: August 27, 1998, as supplemented

by letter dated October 1, 1998.

Brief description of amendment: This amendment revises Technical

Specifications (TS) 3.0.4 and 4.0.4 in accordance with the guidance

provided in Generic Letter 87-09. The revision to TS 3.0.4 removes the

need to explicitly reference its applicability for certain TS. As a

result, several other TS were also amended by deleting references to TS

3.0.4.

Date of issuance: October 20, 1998.

Effective date: October 20, 1998.

Amendment No: 84.

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

Technical Specifications.

Date of initial notice in Federal Register: September 8, 1998 (63

FR 47529).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 20, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

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

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1

and 2, Will County, Illinois

Date of application for amendments: August 23, 1996.

Brief description of amendments: The amendments revise the

Technical Specifications related to the Non-Accessible Area Exhaust

Filter Plenum Ventilation System to reflect the design lineup and to

make provisions for the performance of maintenance and testing.

Date of issuance: October 15, 1998.

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

Amendment Nos.: 105; 105 & 97; 97.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

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

11488).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 15, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: October 16, 1996, as supplemented by

letters dated December 22, 1997, and May 27, 1998.

Brief description of amendment: The amendment changes the Appendix

A Technical Specifications by relocating certain administrative

controls to Quality Assurance Program Manual as described in

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

Controls related to Quality Assurance;'' changing shift coverage from

8-hour day, 40-hour weeks to an option of 8 or 12 hour days and nominal

40-hour weeks; and making editorial changes to the titles of certain

organizational positions.

Date of issuance: October 19, 1998.

Effective date: October 19, 1998, to be implemented within 60 days.

Amendment No.: 146.

Facility Operating License No. NPF-38: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 9, 1997 (62 FR

17233).

The December 22, 1997, and May 27, 1998 letters, provided

clarifying information that did not change the initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 19, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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 application for amendment: June 21, 1995.

Brief description of amendment: The amendments revise the Technical

Specification action statements and certain surveillances of TS 3/

4.5.1, Safety Injection Tanks (SITs). These revisions include a two-

tiered extension of the action completion/allowed outage time for the

SITs. The revisions are also consistent with the guidance provided in

Generic Letter 93-05, ``Line-Item Technical Specifications Improvements

to Reduce surveillance requirements for Testing During Power

Operation.''

Date of Issuance: October 16, 1998.

Effective Date: To be implemented within 30 days from date of

receipt.

Amendment Nos.: 157 and 96.

Facility Operating License No. NPF-16: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 27, 1995 (60

FR 49936).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 16, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Community College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.

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

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

Date of amendment request: October 31, 1996, supplemented October

31, 1997, May 27, 1998, and September 25, 1998.

[[Page 59600]]

Description of amendment request: The amendments revise the

administrative control specifications to reduce the administrative

burden carried by the Facility Review Group and the Plant General

Manager by making more efficient use of site personnel possessing the

requisite experience and qualifications in the review and approval

process for plant procedures.

Date of Issuance: October 16, 1998.

Effective Date: October 16, 1998.

Amendment Nos.: 158 and 97.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the Technical Specifications.

Date of Initial Notice in Federal Register: December 18, 1996 (61

FR 66707) The October 31, 1997, May 27, 1998, and September 25, 1998,

submittals provided clarifying information that did not change the

original no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 16, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Community College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.

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

Generating Station, Ocean County, New Jersey

Date of application for amendment: August 21, 1998.

Brief description of amendment: The amendment removes the

requirement for the Automatic Depressurization System function of the

Electromatic Relief Valves to be operable during Reactor Vessel

Pressure Testing. Additionally, it clarifies Note h of Technical

Specification Table 3.1.1.

Date of Issuance: October 14, 1998.

Effective date: October 14, 1998, to be implemented within 30 days.

Amendment No.: 199.

Facility Operating License No. DPR-16: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 10, 1998 (63

FR 48527).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated October 14, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

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

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

Generating Station, Ocean County, New Jersey

Date of application for amendment: May 28,1998.

Brief description of amendment: The amendment revises Technical

Specification 4.5.A.1 such that the first Type A test required by the

primary containment leakage rate testing program be performed during

refueling outage 18 rather than refueling outage 17.

Date of Issuance: October 15, 1998.

Effective date: October 15, 1998, to be implemented within 30 days.

Amendment No.: 200.

Facility Operating License No. DPR-16. Amendment revised the

Technical Specifications.

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

38201).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated October 15, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

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

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit

1, DeWitt County, Illinois.

Date of application for amendment: May 4, 1998, as supplemented

September 23, 1998.

Brief description of amendment: The amendment incorporates

Technical Specification requirements for the protection systems for the

new static VAR compensators.

Date of issuance: October 9, 1998.

Effective date: October 9, 1998.

Amendment No.: 117.

Facility Operating License No. NPF-62: The amendment revised the

Technical Specifications.

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

30264).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 9, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, IL 61727.

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

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

Date of application for amendment: May 9, 1997, as supplemented

August 4, 1998.

Brief description of amendment: The amendment revises the shutdown

margin requirements and adds Technical Specification 3/4.3.5 to provide

the limiting condition for operation and surveillance requirements for

the shutdown margin monitors. The amendment also makes administrative

changes and revises the associated Bases section.

Date of issuance: October 21, 1998.

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

within 60 days from the date of issuance.

Amendment No.: 164.

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

Facility Operating License and the Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33129).

The August 4, 1998, letter provided clarifying information that did

not change the scope of the May 9, 1997, 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 October 21, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,

Unit No. 1, Washington County, Nebraska

Date of amendment request: July 11, 1995.

Brief description of amendment: The amendment revises Technical

Specifications (TS) 2.3(2)f and 2.3(2)g to increase allowed outage

times for the safety injection tanks (SIT).

Date of issuance: October 19, 1998.

Effective date: October 19, 1998.

Amendment No.: 186.

Facility Operating License No. DPR-40. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 2, 1995 (60 FR

39447). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 19, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102.

[[Page 59601]]

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,

Unit No. 1, Washington County, Nebraska

Date of amendment request: October 3, 1997, as supplemented by

letter dated May 18, 1998.

Brief description of amendment: The amendment revises Technical

Specifications (TS) 3.9 to clarify required flow paths for testing the

auxiliary feedwater system (AFW) and to delete specific AFW pump

discharge pressure.

Date of issuance: October 19, 1998.

Effective date: October 19, 1998, to be implemented 30 days from

the date of issuance.

Amendment No.: 187.

Facility Operating License No. DPR-40: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 3, 1997 (62 FR

63982).

The May 18, 1998, supplemental letter provided additional

clarifying information that did not change the staff's original no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated October 19, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102.

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

Power and Light Company, and Atlantic City Electric Company, Docket No.

50-277, Peach Bottom Atomic Power Station, Unit No. 2, York County,

Pennsylvania

Date of application for amendment: July 10, 1998, as supplemented

by two letters dated September 11, 1998. The supplemental letters

provided clarifying information but did not change the initial no

significant hazards consideration determination.

Brief description of amendment: This amendment revises the

Technical Specifications for safety limit Minimum Critical Power Ratio

from its current value of 1.11 to 1.10 for two recirculation loop

operation, and from 1.13 to 1.12 for single recirculation loop

operation.

Date of issuance: October 26, 1998.

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

startup for Cycle 13 operations, scheduled for October 1998.

Amendment No.: 226.

Facility Operating License No. DPR-44: Amendment revised the

Technical Specifications.

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

FR 48261).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 26, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Publications

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

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments: March 16, 1998, as supplemented

by letters dated May 22, August 10, and September 17, 1998, and also by

letter dated February 9, 1998.

Brief description of amendments: The amendment authorized changes

to the Final Safety Analysis Report to incorporate the increases in the

main steam line radiation monitor setpoint and allowable values and the

change to the design basis of the offgas system to a detonation

resistant design.

Date of issuance: October 13, 1998.

Effective date: October 13, 1998.

Amendment Nos.: 179 and 152.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments

revised the Final Safety Analysis Report.

Date of initial notice in Federal Register: May 20, 1998 (63 FR

27764).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 13, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments: April 23, 1998.

Brief description of amendments: These amendments change the name

``Pennsylvania Power & Light Company'' to ``PP&L, Inc.'' in the

operating licenses and appendices to reflect the licensee's corporate

name change.

Date of issuance: October 19, 1998.

Effective date: Both units, as of the date of issuance to be

implemented within 30 days.

Amendment Nos.: 180 and 153.

Facility Operating License Nos. NPF-14 and NPF-22. The amendments

revised the operating licenses and Appendix B to each licensee and

Attachment 1 to the Unit 1 license.

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

35993).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 19, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendments: February 25, 1997, as

supplemented September 8 and November 18, 1997 and January 8 and July

2, 1998. The supplemental letters provided clarifying information and

did not change the initial proposed no significant hazards

consideration determination.

Brief description of amendments: These amendments revise the

Facility Operating Licenses, Technical Specifications, and

Environmental Protection Plans to reflect a corporate name change,

remove obsolete information, and correct typographical errors.

Date of issuance: October 23, 1998.

Effective date: Both units, as of date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 131 and 92.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications and Licenses.

Date of initial notice in Federal Register: June 4, 1997 (62 FR

30642).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

[[Page 59602]]

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

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

Date of application for amendment: June 25, 1997, as supplemented

August 3, 1998.

Brief description of amendment: The amendment allows the use of

zirconium or stainless steel filler rods in fuel assemblies to replace

failed or damaged fuel rods.

Date of issuance: October 8, 1998.

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

30 days.

Amendment No.: 183.

Facility Operating License No. DPR-64: Amendment revised the

Technical Specifications.

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

33107).

The August 3, 1998, submittal fell within the scope of, and did not

change, the initial proposed finding of no significant hazards

consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 8, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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: August 8, 1997, as supplemented

by letters dated March 9, May 6, July 6, July 31, September 4, and

September 11, 1998.

Brief description of amendments: The amendments revise the

Technical Specifications to accommodate an increase in the maximum

licensed thermal power level from 2558 megawatts thermal (MWt) to 2763

MWt.

Date of issuance: October 22, 1998.

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

Unit 1 prior to startup from the next refueling outage and on Unit 2

prior to startup from the current refueling outage.

Amendment Nos.: Unit 1-214; Unit 2-155.

Facility Operating License Nos. DPR-57 and NPF-5: The amendments

revised the Technical Specifications and Operating Licenses.

Public comments requested as to proposed no significant hazards

consideration: Yes. (63 FR 53730 dated October 6, 1998.) The notice

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

no significant hazards consideration determination. No comments have

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

hearing by November 5, 1998, but indicated that if the Commission makes

a final no significant hazards consideration determination, any such

hearing would take place after issuance of the amendments.

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

exigent circumstances, and a final no significant hazards consideration

determination are contained in a Safety Evaluation dated October 22,

1998.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia.

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

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

Alabama

Date of amendments request: May 27, 1997.

Brief Description of amendments: The amendments revise the

Technical Specifications (TSs) to change the Applicable Modes for

Source Range (SR) Nuclear Instrumentation (NI) (TS \3/4\.3.1, ``Reactor

Trip System Instrumentation''), provide allowances for an exception to

the requirements for the state of the power supplies for residual heat

removal discharge to charging pump suction valves following Mode

changes (TS \3/4\.5.2, ``ECCS Subsystems--Tavg>350 deg.F''

and \3/4\.5.3, ``ECCS Subsystems--Tavg<350 deg.F''), and

delete cycle-specific guidance concerning manual engineered safety

feature functional input checks.

Date of issuance: October 15, 1998.

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

30 days from the date of issuance.

Amendment Nos.: Unit 1-138; Unit 2-130.

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

the Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33134).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 15, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

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

Unit 1, Rhea County, Tennessee

Date of application for amendment: June 5, 1997, as supplemented

April 21 and August 12, 1998.

Brief description of amendment: The requested changes would revise

the Technical Specifications (TS) to allow testing of diesel

generators, pursuant to Surveillance Requirement (SR) 3.8.1.14, during

operational modes 1 or 2. The requested changes would also revise the

TS to allow testing of the diesel generator batteries and associated

battery chargers, pursuant to SRs 3.8.4.12, 3.8.4.13 and 3.8.4.14

during operational modes 1, 2, 3 or 4.

Date of issuance: October 19, 1998.

Effective date: October 19, 1998.

Amendment No.: 12.

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

Date of initial notice in Federal Register: July 29, 1998 (63 FR

40561).

The supplemental letter dated August 12, 1998, contained clarifying

information and did not change the original no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 19, 1998.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: April 8, 1998, as revised by

letter dated August 27, 1998.

Brief description of amendment: The amendment reduces the allowable

reactor coolant system specific activity from 1.0 microcurie/gram to

0.20 microcurie/gram dose equivalent I-131, a means described by

Generic Letter 95-05 to support the reduction of reactor coolant system

specific activity limits.

Date of issuance: October 27, 1998.

Effective date: October 27, 1998.

Amendment No.: 140.

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

Technical Specifications.

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

FR 49137).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 27, 1998.

[[Page 59603]]

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

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

Notice of Issuance of Amendment to Facility Operating License and

Final No Significant Hazards Consideration Determination

During the period since publication of the last biweekly notice,

individual notices of issuance of amendments have been issued for the

facilities as listed below. These notices were previously published as

separate individual notices. They are repeated here because this

biweekly notice lists all amendments that have been issued for which

the Commission has made a final determination that an amendment

involves no significant hazards consideration.

In this case, a prior Notice of Consideration of Issuance of

Amendment, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing was issued, a hearing was requested, and

the amendment was issued before any hearing because the Commission made

a final determination that the amendment involves no significant

hazards consideration.

Details are contained in the individual notice as cited.

Notice of Issuance of Amendments to Facility Operating Licenses and

Final Determination of No Significant Hazards Consideration and

Opportunity for a Hearing (Exigent Public Announcement or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

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

and regulations. The Commission has made appropriate findings as

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

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

Because of exigent or emergency circumstances associated with the

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

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

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

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

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

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

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

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

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

transcribed as appropriate and the licensee has been informed of the

public comments.

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

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

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

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

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

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

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

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

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

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

significant hazards consideration is involved.

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

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

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

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

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

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

available for public inspection at the Commission's Public Document

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

the local public document room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By December 4, 1998, 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

[[Page 59604]]

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

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

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

effect.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

Arizona Public Service Company, et al., Docket No. STN 50-530, Palo

Verde Nuclear Generating Station, Unit No. 3, Maricopa County, Arizona

Date of application for amendment: October 6, 1998

Brief description of amendment: The amendment revises TS 3.3.1,

``Reactor Protective System (RPS) Instrumentation--Operation,'' and TS

3.3.2, ``Reactor Protective System (RPS) Instrumentation--Shutdown.''

The proposed amendment would clarify the power level threshold at which

certain RPS instrumentation trips must be enabled and may be bypassed,

and would clarify that this level is a percentage of the neutron flux

at rated thermal power (RTP). The bypass power level, 1E-4% RTP, would

be specified as logarithmic power instead of thermal power.

Date of issuance: October 19, 1998.

Effective date: October 19, 1998.

Amendment No.: 119.

Facility Operating License No. NPF-74: The amendment revised the

Technical Specifications.

Press release issued requesting comments as to proposed no

significant hazards consideration: Yes. October 13, 1998. Arizona

Republic Newspaper (Arizona).

Comments received: No. The Commission's related evaluation of the

amendment, finding of exigent circumstances, consultation with the

State of Arizona and final determination of no significant hazards

consideration are contained in a Safety Evaluation dated October 19,

1998.

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 Project Director: William H. Bateman.

Dated at Rockville, Maryland, this 28th day of October 1998.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

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

Nuclear Reactor Regulation.

[FR Doc. 98-29433 Filed 11-3-98; 8:45 am]

BILLING CODE 7590-01-P