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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from December 18, 1997, through January 2, 1998.

The last biweekly notice was published on December 31, 1997 (62 FR

68303).

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 Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and should cite the publication date and

page number of this Federal Register

[[Page 2272]]

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 February 13, 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 Public

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

1, 2, and 3, Maricopa County, Arizona

Date of amendments request: December 17, 1997.

Description of amendments request: The proposed amendment would

modify the Technical Specifications (TS) to replace the current

explicit reference to Exide batteries with a generic reference to low

specific gravity cells. The proposed change would also remove footnotes

for the Unit 2 and Unit 3 TS that referred to one time exemptions that

no longer apply. The proposed change would allow replacement of the

existing Class 1E, 125 volt DC batteries with equivalent batteries

manufactured by different vendors.

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.

[[Page 2273]]

The Class 1E 125V DC system provides DC power to the Class 1E DC

loads for operation, control and switching, including the inverters

which power the Class 1E 120V vital AC busses. This system is not an

accident initiator. It is, however, an accident mitigation system.

The replacement low specific gravity rectangular cell batteries have

been designed to IEEE 485-1978 standards and meet all appropriate

seismic criteria. There is no change in the physical or electrical

separation provisions for the Class 1E 125V DC channels. These

batteries are used extensively throughout the industry and their

failure mechanisms are well understood. The existing high specific

gravity round cell batteries are experiencing premature capacity

loss for which a definitive root cause of failure has not been

determined. Therefore, replacement of the high specific gravity

round cell batteries with low specific gravity rectangular cell

batteries increases the overall reliability of the Class 1E 125 V DC

system. In addition, the design requirements of the replacement

batteries ensures that the batteries will be capable of reliably

performing their design function during all modes of operation and

will serve to mitigate any accident that may occur. The proposed

amendment does not change the performance criteria or cell

parameters for the Class 1E 125V DC sources that are defined in the

current Technical Specifications for each unit. Since this change is

increasing the overall reliability and performance of the system and

is designed to meet the same stringent requirements of the existing

high specific gravity round cell batteries, it does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Replacement of the high specific gravity round cell batteries

will occur during acceptable modes of operation as defined in the

current Technical Specifications for each unit, i.e., the work will

be performed during Modes 5 or 6, or with the reactor defueled.

Technical Specification 3.8.2.2, DC Sources--Shutdown, for each unit

requires one Class 1E 125V DC train to be operable in Modes 5 or 6.

With one Class 1E 125V DC train operable, the other train may be

removed from service for battery cell replacement. Since the battery

cell replacement will be performed within the Limiting Condition of

Operation for DC Sources--Shutdown, the replacement sequence of the

battery banks will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

Low specific gravity rectangular cell batteries have been used

throughout the industry for many years. Some cells have been in

service for 17 years and have not degraded to the extent that they

require replacement. The low specific gravity rectangular cell

batteries have demonstrated good reliability and the failure

mechanisms associated with these batteries are well understood. The

high specific gravity round cell batteries that are currently

installed are exhibiting premature capacity loss for which a

definitive root cause of failure has not been determined. Replacing

the high specific gravity round cell batteries with low specific

gravity rectangular cell batteries that have seen extensive use in

the industry, are well understood and have been designed to meet the

same stringent requirements as that of the existing batteries

ensures that the overall system reliability is increased. No new or

common mode failures are created since the replacement low specific

gravity rectangular cell batteries have been designed to the same

stringent requirements as the existing batteries. The proposed

amendment does not change the performance criteria or cell

parameters for the Class 1E 125V DC sources that are defined in the

current Technical Specifications for each unit. Therefore,

replacement of the high specific gravity round cell batteries with

low specific gravity rectangular cell batteries 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.

As described previously, replacing the high specific gravity

round cell batteries with low specific gravity rectangular cell

batteries enhances the overall system reliability. The low specific

gravity rectangular cell batteries have been designed to the same

criteria as the existing high specific gravity round cell batteries.

The performance criteria and cell parameters specified in each

unit's Technical Specifications are not affected by this change.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

Replacement of the high specific gravity round cell batteries

will occur during acceptable modes of operation as defined in the

current Technical Specifications for each unit, i.e., the work will

be performed during Modes 5 or 6, or with the reactor defueled.

Technical Specification 3.8.2.2, DC Sources--Shutdown, for each unit

requires one Class 1E 125V DC train to be operable in Modes 5 or 6.

With one Class 1E 125V DC train operable, the other train may be

removed from service for battery cell replacement. Since the battery

cell replacement will be performed within the Limiting Condition of

Operation for DC Sources--Shutdown, the work sequence for

replacement of the battery banks will not involve a significant

reduction in a margin of safety.

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

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

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

amendments request involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

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

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

Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: William H. Bateman.

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

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

Date of amendment request: October 3, 1996.

Description of amendment request: The proposed amendments would

correct a typographical error which was introduced into the Technical

Specifications (TS) with issuance of Amendment Nos. 150 and 145.

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 because of the

following:

The proposed change does not alter the manner of operation of

the facility, it merely restores the correspondence between the

applicability of the Limiting Conditions for Operability (LCO) for

the (Source Range) Neutron Monitors and the Drywell Radiation

Monitors and the associated Surveillance Requirements for the same

two instrument functions as described in Tables 3.2.F-1 and 4.2.F-1.

No changes are proposed which will affect the probability of an

accident previously evaluated, since the instruments and their

associated functions are credited to operate during and after a

postulated accident. The function of a device after an event has

occurred cannot affect the probability of that accident occurring.

Similarly, the proposed changes do not effect the operation or

function of structures, systems or components which effect the

probability of any accident previously evaluated.

The proposed changes do not affect the consequence of an

accident previously evaluated since the changes do not decrease the

availability of any functions credited with performing mitigative

actions. The availability requirements of the Drywell Radiation

Monitors is not changed because the associated LCO requires the

monitors to be OPERABLE in the conditions proposed in this change.

The proposed change merely assures that the surveillance

requirements are met in the modes which correspond to the LCO. The

(Source Range) Neutron Monitor surveillance requirements change does

not affect the ability of the system to provide adequate information

to the operators to mitigate the consequences of a postulated

accident, since the system OPERABILITY requirements as specified in

the LCO are not affected.

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

accident from any accident previously evaluated because:

[[Page 2274]]

The proposed change does not introduce any new or different

types of operation of the plants. No new equipment is introduced as

a result of the implementation of the proposed change. Therefore no

changes are proposed which could introduce a new or different kind

of accident from any previously evaluated.

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

because:

The proposed change does not effect the margin of safety. The

LCO requirements for the two instrument systems which are effected

are not changed; the OPERABILITY requirements remain the same. The

only substantive changes are the modes in which surveillance testing

is required to be performed. The change restores the need to perform

testing of the Drywell Radiation Monitor prior to and during

OPERATIONAL MODE 3 operations, and removes the requirement to

perform testing of the (Source Range) Neutron Monitors prior to and

during operation in MODE 3 when it is not required to be OPERABLE as

described in the associated LCO. Based on this, the availability of

the affected instruments to perform their design function is not

effected by this change and no reduction in the margin of safety is

proposed.

Guidance has been provided in ``Final Procedures and Standards

on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,

for the application of standards to license change requests for

determination of the existence of significant hazards

considerations. This document provides examples of amendments which

are and are not considered likely to involve significant hazards

considerations.

This proposed amendment does not involve any irreversible

changes, significant relaxation of the criteria used to establish

safety limits, a significant relaxation of the bases for the

limiting safety system settings, or a significant relaxation of the

bases for the limiting conditions for operations. Therefore, based

on the guidance provided in the Federal Register and the criteria

established in 10 CFR 50.92(c), the proposed change does not

constitute 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

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

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

NRC Project Director: Robert A. Capra.

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: August 29, 1997.

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 Siemens' methodologies that will enhance

operational flexibility and reduce the likelihood of future plant

derates, (b) administrative changes that both eliminate the cycle

specific implementation of Atrium-9B fuel, and (c) changes to the

Dresden and Quad Cities Minimum Critical Power Ratio (MCPR).

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.

Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1

and 2)

The Reference 1 [ANF-91-048(P), Supplement 1, ``BWR Jet Pump

Model Revision for RELAX''. Submitted to the NRC by SPC letter, ANF-

91-048(P), Supplement 1 and ANF-91-048(NP), Supplement 1, ``BWR Jet

Pump Model Revision for RELAX,'' RAC:96-042, R.A. Copeland to US

NRC, May 6, 1996] 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.

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 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] 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.

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

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

Uncertainty For Limited Data Sets''. Submitted to the NRC by SPC

letter, ``Request for Review of ANFB Critical Power Correlation

Uncertainty for Limited Data Sets, ANF-1125(P), Supplement 1,

Appendix D'', HDC:97:032, H.D. Curet to Document Control Desk, April

18, 1997] 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.

[[Page 2275]]

Reference 7 documents the additive constant uncertainty for 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 more than 99.9% 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 of consequences of an accident previously evaluated are

not affected by these changes.

Change to Minimum Critical Power Ratio Safety Limit (Quad Cities

Units 1 and 2 and Dresden Units 2 and 3)

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

Dresden Units 2 and 3 will not increase the probability of an

accident previously evaluated. This change implements the MCPR

Safety Limits resulting from the SPC ANFB critical power correlation

methodology using a revised additive constant uncertainty from

Reference 7. The MCPR Safety Limit of 1.09 that is proposed for Quad

Cities Units 1 and 2 and Dresden Units 2 and 3 is 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. Changing

the MCPR Safety Limit will not alter any physical systems or

operating procedures. The MCPR Safety Limit is set to 1.09, which is

the CPR value where less than 0.1% of the rods in the core are

expected to experience boiling transition. This safety limit is

expected to be applicable for future cycles of ATRIUM-9B at Dresden

and Quad Cities. Therefore the probability or consequences of an

accident will not increase.

Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel

Reloads (Quad Cities Unit 2 and Dresden Unit 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 is not expected to increase.

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 average planar linear

heat generation rate] limits and their exposure basis to be

specified in the COLR [core operating limits report] and to

establish consistency with approved methodologies currently utilized

by Siemens Power Corporation, which calculates 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 [boiling water reactor]. 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

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

because the exposure basis of the APLHGR 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.

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.

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.

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. Therefore, no

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

kinds of accidents are created.

Change to Minimum Critical Power Ratio Safety Limit (Quad Cities

Units 1 and 2 and Dresden Units 2 and 3)

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%

of the rods avoid boiling transition.

The MCPR Safety Limit is changing for Quad Cities Unit 1 due to

the transition to SPC ATRIUM-9B fuel and SPC methodologies. The MCPR

Safety Limit is changing for Quad Cities Unit 2 due to the Reference

7 methodology, which documents a 0.0195 ATRIUM-9B additive constant

uncertainty and supports a 1.09 MCPR Safety Limit. This MCPR Safety

Limit is lower than the current MCPR Safety Limit for Quad Cities

Unit 2, 1.10, which is based on a higher interim conservative

additive constant uncertainty of 0.029. The lower ATRIUM-9B additive

constant uncertainty results in the lower MCPR Safety Limit for Quad

Cities Unit 2. The new MCPR Safety Limit for Dresden Units 2 and 3,

1.09, is greater than the current value at Dresden Units 2 and 3 and

is being increased now in anticipation of bounding future reloads of

ATRIUM-9B. Therefore, no new accidents are created that are

different from any accident previously evaluated.

[[Page 2276]]

Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel

Reloads (Quad Cities Unit 2 and Dresden Unit 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 the generic Reference 3 methodology

and the anticipated approval of the Reference 7 additive constant

uncertainty methodology, these footnotes are no longer applicable.

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.

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.

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

the following reasons:

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 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.

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 this reference

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. 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 topical 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-95-021(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] that greater than 99.9% 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.

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% of the rods are expected to be in boiling

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

Specification amendment 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 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 only ensure

that the appropriate level of fuel protection is being employed.

Change to Minimum Critical Power Ratio Safety Limit (Quad Cities

Unit 1 and 2 and Dresden Units 2 and 3)

Changing the MCPR Safety Limit for Quad Cities and Dresden 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% of the

rods are expected to be in boiling transition if the MCPR Safety

Limit is not violated. The proposed Technical Specification

amendment reflects the MCPR Safety Limit results from conservative

evaluations by SPC using the ANFB critical power correlation with

the new 0.0195 ATRIUM-9B additive constant uncertainty documented 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%

of the fuel rods avoiding transition boiling during normal operation

as well as during an anticipated operational occurrence is met.

Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel

Reloads (Quad Cities Unit 2 and Dresden Unit 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) have been submitted to the NRC and both are proposed

to be added to the Technical Specifications in this amendment, 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.

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 La Salle, Jacobs Memorial Library, Illinois Valley Community

College, Oglesby, Illinois 61348; for Quad Cities, Dixon Public

Library, 221 Hennepin Avenue, Dixon, Illinois 61021.

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

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

NRC Project Director: Robert A. Capra.

[[Page 2277]]

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

Date of amendment request: October 27, 1997.

Description of amendment request: The proposed amendments would

clarify the applicability, action and surveillance requirements for the

Standby Liquid Control System.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because of the

following:

The proposed changes represent the conversion of current

requirements which are based on generic guidance or previously

approved provisions for other stations. The proposed changes are

consistent with NUREG-1433 and do not significantly increase the

probability or consequences of any previously evaluated accidents

for Dresden or Quad Cities Stations. The proposed amendment is

consistent with the current safety analyses and represents

sufficient requirements for the assurance and reliability of

equipment assumed to operate in the safety analysis, or provide

continued assurance that specified parameters remain within their

acceptance limits. The proposed TS continue to ensure sufficient

requirements are in place for the SLCS during plant operation. The

proposed changes that eliminate Applicability and Actions during

refueling operations for the SLCS do not affect the probability of

any previously evaluated accident because only one control rod can

be withdrawn during refueling operations and Shutdown Margin

requirements are maintained in the Technical Specifications.

Therefore, the probability of an inadvertent criticality is not

increased as reactivity controls are maintained. Because the SLCS is

manually initiated and not assumed to mitigate any accident scenario

during refueling operations, the proposed changes do not affect the

consequences of any previously evaluated accident. As such, these

changes will not significantly increase the probability or

consequences of a previously evaluated accident.

The associated systems related to this proposed amendment are

not assumed in any safety analysis to initiate any accident sequence

for Dresden or Quad Cities Stations. In addition, the revisions

proposed to the surveillance requirements are administrative in

nature and either relocate procedural details to administrative

controls or allow provisions for manual alignment of a manual system

to the proper orientation. As such, because there is no effect on

any accident scenario, the probability of any accident previously

evaluated is not increased by the proposed amendment. Because the

proposed changes are administrative in nature, the consequences of

any previously evaluated accident are not increased.

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

accident from any accident previously evaluated because:

The proposed amendment for Dresden and Quad Cities Station's

Technical Specification is based on generic guidance or NRC accepted

changes for later operating BWR plants. The proposed amendment has

been reviewed for acceptability at the Dresden and Quad Cities

Nuclear Power Stations considering similarity of system or component

design versus the generic guidance. The proposed changes do not

create the possibility of a new or different kind of accident

previously evaluated for Dresden or Quad Cities Stations. No new

modes of operation are introduced by the proposed changes. SLCS

requirements are adequately retained to ensure sufficient controls

remain during plant operations. The proposed changes to the

Applicability and Actions during refueling operations for the SLCS

do not create a new or different kind of previously evaluated

accident. Because the SLCS is manually initiated to mitigate

accident concerns during power operations, the proposed deletion of

Applicability and Actions during refueling operations does not

affect the probability of a new or different kind of accident from

being created. The changes proposed to the surveillance requirements

are administrative in nature and do not affect the system operation;

as such, the proposed changes do not affect the probability of a new

or different kind of accident being created. Therefore, the proposed

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

accident from any previously evaluated.

The associated systems related to this proposed amendment are

not assumed in any safety analysis to initiate any accident sequence

for Dresden or Quad Cities Stations; therefore, the proposed changes

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

from any previously evaluated.

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

because:

The proposed amendment represents the conversion of current

requirements which are based on generic guidance or previously

approved provisions for other stations. The proposed changes are

consistent with NUREG-1433 and do not adversely affect existing

plant safety margins or the reliability of the equipment assumed to

operate in the safety analysis. The proposed changes have been

evaluated and found to be acceptable for use at Dresden or Quad

Cities based on system design, safety analysis requirements and

operational performance. SLCS provisions continue to be adequately

maintained during plant operation. The proposed changes to the

Applicability and Actions during refueling operations for the SLCS

do not significantly reduce existing plant safety margins. Because

the SLCS is manually initiated to mitigate accident concerns during

power operations, the proposed deletion of Applicability and Actions

during refueling operations has no effect on existing plant safety

margins as this system is not required during this mode of

operation. The changes proposed to the surveillance requirements are

administrative in nature and do not affect the system operation; as

such, the proposed changes do not adversely affect existing plant

safety margins as adequate system surveillance requirements are

maintained. Since the proposed changes are based on NRC accepted

provisions at other operating plants that are applicable at Dresden

or Quad Cities and maintain necessary levels of system or component

reliability, the proposed changes do not involve a significant

reduction in the margin of safety.

The proposed amendment for Dresden and Quad Cities Stations will

not reduce the availability of systems required to mitigate accident

conditions; therefore, the proposed changes do not involve a

significant reduction in the margin of safety.

Guidance has been provided in ``Final Procedures and Standards

on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,

for the application of standards to license change requests for

determination of the existence of significant hazards

considerations. This document provides examples of amendments which

are and are not considered likely to involve significant hazards

considerations.

This proposed amendment does not involve a significant

relaxation of the criteria used to establish safety limits, a

significant relaxation of the bases for the limiting safety system

settings or a significant relaxation of the bases for the limiting

conditions for operations. Therefore, based on the guidance provided

in the Federal Register and the criteria established in 10 CFR

50.92(c), the proposed change does not constitute 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

requested amendments involve 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.

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

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

NRC Project Director: Robert A. Capra.

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

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

Date of amendment request: November 7, 1997.

Description of amendment request: The proposed amendments would

[[Page 2278]]

relocate the Unit 2 24/48 Vdc batteries, chargers, and distribution

systems operability and surveillances requirements from the Technical

Specifications to licensee administratively controlled documents.

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 because of the

following:

Removal of the Unit 2 24/48 Vdc battery, charger, and

distribution panel requirements from the Technical Specification

requirements of 3/4.9.C, 3/4.9.D, 3/4.9.E, and 3/4.9.F and the

subsequent relocation of those requirements to licensee

administrative controls is an administrative change that will

continue [to] ensure the availability of the Unit 2 24/48 Vdc system

and will not increase the probability of accidents previously

evaluated. Relocation of the Unit 2 24/48 Vdc requirements to

administrative controls will have no effect on the control

instrumentation and cannot act as an initiator for any of the

accidents evaluated in the UFSAR [Updated Final Safety Analysis

Report].

Similarly, relocation of the Unit 2 24/48 Vdc system

requirements to licensee administrative controls will have no effect

on the availability of the loads which are supplied by the Unit 2

24/48 Vdc batteries nor on any of the consequences of accidents

previously evaluated in the UFSAR. Control of the Unit 2 24/48 Vdc

requirements by licensee administrative controls under 10 CFR 50.59

will not affect any of the protection or mitigation functions which

may be provided by any of the loads supplied by the batteries.

Operation under the proposed amendment will not significantly

increase the probability or consequences of any accidents previously

evaluated.

Because of the above evaluation, removal of the Unit 2 24/48 Vdc

system from the Technical Specifications will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

accident from any accident previously evaluated because:

The Unit 2 24/48 Vdc batteries, chargers, and other components

will retain the separation, and redundancy under which they are

presently installed. No new failure modes are introduced by this

administrative relocation of requirements, for the Unit 2 24/48 Vdc

system, from the Technical Specifications to licensee administrative

control. Since the batteries are not being operated differently and

transferring ATS [Analog Trip System] loads to the 125 Vdc safety-

related battery system does not affect the function or mode of

operation of these loads, the possibility of a new or different

accident from any accident previously evaluated is not increased or

created by this administrative change.

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

because:

Relocation of the TS requirements for the Unit 2 24/48 Vdc

system does not affect the operating points or setpoints of any

systems or components. Plant operating points or parameters are not

changed by the proposed relocation of requirements in this amendment

request. The safety-related equipment that is supported by the Unit

2 24/48 Vdc system will continue to be required in the existing

modes of applicability as determined by the individual equipment

Technical Specifications. Thus operation under the proposed license

amendment removes some redundancy and constraints during refueling

but does not significantly reduce the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

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

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

NRC Project Director: Robert A. Capra.

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

Unit 1, LaSalle County, Illinois

Date of amendment request: November 24, 1997.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.3.2, ``Isolation Actuation

Instrumentation,'' to add instrumentation for the reactor water cleanup

(RWCU) pump rooms and valve room as a result of modifications to the

RWCU system. Also, additional instrumentation will be added in the RWCU

holdup pipe area, the filter/demineralizer valve rooms, and RWCU pump

suction high flow switch as a result of a high energy line break re-

evaluation. The setpoints for the RWCU heat exchanger room

instrumentation will be revised as a result of new design basis

calculations. The proposed amendment will also delete instrumentation

related to the residual heat removal (RHR) steam condensing mode which

is no longer utilized and will eliminate the alarm and isolation

functions for the RHR shutdown cooling mode.

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 because:

(a) There is no effect on accident initiators so there is no

change in [the] probability of an accident. A line break in the

subject areas would consist of an instantaneous circumferential

break downstream of the outermost isolation valve of one of these

systems. The leak detection isolation is only a precursor of a

break, and thus does not affect the probability of a break.

(b) There is no or minimal effect on the consequences of

analyzed accidents, due to changing the leak detection ambient T or

Delta T setpoint and allowable values to detect 25 gpm equivalent

leakage. The addition of more ambient T and Delta T leak detection

monitoring, along with the addition of the high flow break detection

will actually decrease the consequences of the associated accidents.

The worst case accident outside the primary containment boundary is

a main steam line break which bounds the dose consequences of all

line breaks and therefore bounds any size of leak.

The deletion of the RHR steam condensing mode isolation

actuation instrumentation trip functions from the LaSalle TS does

not increase the probability or consequences of an accident

previously evaluated, because this mode of operation of the RHR

system has been deleted from the LaSalle design basis and the lines

that were previously high energy line are isolated during unit

operation, including Operational Condition 1 (Run mode), Operational

Condition 2 (Startup mode), and Operational Condition [3] (Hot

Shutdown).

The deletion of the RHR shutdown cooling mode leak detection T

and Delta T isolation actuation instrumentation trip functions from

the LaSalle TS does not increase the probability or consequences of

an accident previously evaluated, because the leak detection is only

a precursor of a break, and thus does not affect the probability of

a break. Also, there are two remaining different methods of

detecting abnormal leakage and isolating the system in technical

specification trip functions A.6.a, Reactor Vessel Water Level--Low,

Level 3 and A.6.c, RHR Pump Suction Flow--High. In addition, other

means to detect leakage from the RHR system, such as sump monitoring

and area radiation monitoring, are also available. In accordance

with TS Administrative Requirement 6.2.F.1, LaSalle has a leakage

reduction program to reduce leakage from those portions of systems

outside primary containment that contain radioactive fluids. RHR,

including piping and components associated with the shutdown cooling

mode, is part of this program, which includes periodic visual

inspection for system leakage. The sump monitoring, radiation

monitoring and periodic inspections for system leakage makes the

probability of a leak of 5 gpm going undetected for more than a day

very low.

Also, due to the low reactor pressures (less than 135 psig) at

which RHR shutdown

[[Page 2279]]

cooling mode is able to operate, reactor coolant makeup and outflow

is very low compared to normal plant operation. A change in flow

balance due to a leak is thus more readily detectable with reactor

coolant water level changes and makeup flow rate, and thus precludes

a significant leak going undetected before break detection

instrumentation would cause automatic isolation.

Therefore, there is not a significant increase in the

probability or consequences of an accident previously evaluated.

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

accident from any accident previously evaluated because:

The purpose of the leak detection system, as it applies to the

RWCU and RHR system areas, is to provide the capability for leak

detection and automatic isolation as necessary of the system in the

event of leakage in these areas. This change maintains this

capability with at least two different methods of detection of

abnormal leakage for protection from the flooding concerns of a

significant leak or line break when the RHR system is operating in

the shutdown cooling mode, so that redundant systems will not be

affected.

This change also maintains or adds primary containment isolation

logic for the leak detection isolation based on temperature

monitoring in RWCU areas and break detection based on RWCU pump

suction flow--high. The additional instrumentation and the

associated isolation logic is the same or similar to existing

instrumentation and logic for containment isolation actuation

instrumentation, so no new failure modes are created in this way.

Therefore, the possibility of a new or different kind of

accident from any previously evaluated is not created.

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

because:

The change to the automatic isolation setpoint for high Delta T

leak detection in the heat exchanger rooms is based on current

configuration calculated/analyzed response to a small leak compared

to a circumferential break. The increased leakage rate in the RWCU

heat exchanger rooms that is necessary to actuate isolation on high

temperature during winter conditions, does not adversely affect the

margin of safety. This increased leakage rate is below the critical

crack leakage rate as represented in [Updated Final Safety Analysis

Report] UFSAR Figure 5.2-11. Additionally, differential temperature

leak detection is conservative under these same conditions, and will

actuate isolation at a leakage rate less than the established limit.

The leak detection isolation logic is unchanged and thus remains

single failure proof.

The addition of automatic primary containment isolation on

Ambient and Differential Temperature (Delta T)-High for the Reactor

Water Cleanup System (RWCU) Pump, Pump Valve, Holdup Pipe, and

Filter/Demineralizer (F/D) Valve Rooms and the addition of the RWCU

Pump Suction Flow High line break isolation add to the margin of

safety with respect to leak detection and line breaks in the RWCU

system, because the system isolation diversity is increased and the

amount of system piping monitored for leakage is increased.

The setpoints for the ambient temperature and differential

temperature leak detection isolations being changed or added and the

RWCU pump suction flow--high are set sufficiently high enough so as

not to increase the possibility of spurious actuation. In the event

that a spurious actuation does occur, little safety significance is

presented since the RWCU system performs no safety function. The

setpoints and allowable values for the proposed changes also assure

sufficient margin to the analytical values and [are] high enough to

prevent spurious actuations based on calculations consistent with

Regulatory Guide 1.105.

The deletion of the RHR steam condensing mode isolation

actuation instrumentation does not effect the margin of safety,

because this mode is no longer utilized by LaSalle in Operational

Conditions 1, 2, or 3 (Run mode, Startup Mode, or Hot Shutdown).

The elimination of the temperature based trip functions for the

RHR shutdown cooling mode area is based on the determination that

temperature is not the appropriate parameter as it does not provide

meaningful indication and will not provide setpoints that would be

sufficiently above the normal range of ambient conditions to avoid

spurious isolations.

There are two remaining different methods of detecting abnormal

leakage and isolating the system in technical specification trip

function A.6, namely A.6.a, Reactor Vessel Water Level--Low, Level 3

and A.6.c, RHR Pump Suction Flow--High. In addition, other means to

detect leakage from the RHR system, such as sump monitoring and area

radiation monitoring, are also available. Also, in accordance with

TS Administrative Requirement 6.2.F.1, LaSalle has a leakage

reduction program to reduce leakage from those portions of systems

outside primary containment that contain radioactive fluids. RHR,

including piping and components associated with the shutdown cooling

mode, is part of this program, which includes periodic visual

inspection of system for leakage.

The previous evaluation of diversity of isolation parameters, as

presented in Table 5.2-8 of the UFSAR remains unchanged. Adequate

diversity of isolation parameters is maintained because there are at

least two different methods available to detect and allow isolation

of the system for a line break, as necessary.

Therefore, this requested Technical Specification amendment does

not involve a significant reduction in the margin of safety.

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

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

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

requested amendment involves no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

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

NRC Project Director: Robert A. Capra.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,

Michigan

Date of amendment request: December 10, 1997 (NRC-97-0105).

Description of amendment request: The proposed amendment would

modify the technical specifications (TS) and the bases to accommodate

the installation of an improved power range neutron monitoring system.

The modification and the TS changes represent part of the licensee's

actions in response to Generic Letter 94-02, ``Long-Term Solutions and

Upgrade of Interim Operating Recommendations for Thermal-Hydraulic

Instabilities in Boiling Water Reactors,'' dated July 11, 1994. The TS

revisions include changes to Action Statements and Surveillance

Requirements which are generally consistent with licensing topical

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

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

Trip Function,'' and NEDC-32410P Supplement 1, ``NUMAC PRNM Retrofit

Plus Option III Stability Trip Function,'' which were reviewed by the

NRC as documented in a letter dated September 5, 1995, and a safety

evaluation dated August 15, 1997. The proposed amendment also includes

two unrelated changes. Surveillance Requirement 4.3.1.3 and its

associated bases are modified to clarify the applicability of response

time testing requirements. In addition, the first page of Table 3.3.6-2

is modified to correct a typographical error in the title.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

in the probability or consequences of an accident previously

evaluated.

This proposed TS change is associated with the NUMAC-PRNM

retrofit design. The proposed TS change involves modification of the

Limiting Conditions for Operation (LCOs) and Surveillance

Requirements (SRs) for equipment designed to mitigate events that

result in power increase transients. The APRM [average power range

monitor] system mitigative action is to block control rod withdrawal

or initiate a reactor scram, which terminates the power increase

when setpoints are exceeded. The Rod Block

[[Page 2280]]

Monitor (RBM) system mitigative action is to block continuous

control rod withdrawal prior to exceeding the fuel design limits

during a postulated Rod Withdrawal Error. The functional capability

of the previous Reactor Coolant System Recirculation Flow control

rod block trip functions have been incorporated into the modified

APRM control rod block trip functions. The worst case failure of

either the APRM or the RBM systems is failure to initiate mitigative

action (failure to scram or block rod withdrawal). Failure to

initiate mitigative action will not increase the probability of an

accident. Thus, the proposed change does not increase the

probability of an accident previously evaluated.

For the APRM and the RBM systems, the NUMAC PRNM design,

together with revised operability requirements (LCOs) and revised

testing requirements (SRs), continues to perform the same mitigation

functions under identical conditions with availability comparable to

the types of equipment that it replaces. Because there is no change

in mitigation functions and because availability of the functions is

maintained, the proposed change does not involve a significant

increase in the consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes involve modification and replacement of the

existing power range neutron monitoring equipment, and modification

of the setpoints and operational requirements for the APRM and RBM

systems. These proposed changes do not modify the basic functional

requirements of the affected equipment, create any new system

interfaces or interactions, nor create any new system failure modes

or sequence of events that could lead to an accident. The worst case

failure of the affected equipment is failure to perform a mitigation

action, and failure of this mitigative equipment does not create the

possibility of a new or different kind of accident. The proposed

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

accident from any accident previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed TS change is associated with the NUMAC PRNM

retrofit design. The NUMAC PRNM change does not impact reactor

operating parameters nor the functional requirements of the power

range neutron monitoring system. The replacement equipment continues

to provide information, enforce control rod blocks and initiate

reactor scrams under appropriate specified conditions. The proposed

change does not revise any safety margin requirements. The

replacement APRM/RBM equipment has improved channel trip accuracy

compared to the current system and meets or exceeds system

requirements previously assumed in setpoint analysis. Thus, the

ability of the new equipment to enforce compliance with margins of

safety equals or exceeds the ability of the equipment which it

replaces. The proposed change does not involve a 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. The editorial change in Table 3.3.6-2 and the clarification

in Surveillance Requirement 4.3.1.3 also satisfy the three standards of

10 CFR 50.92(c). Therefore, the NRC staff proposes to determine that

the amendment request involves no significant hazards consideration.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

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: December 1, 1997.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TSs) to add a time delay,

including allowance, to a portion of the Engineered Safety Feature

Actuation System undervoltage (UV) trip TSs. The proposed changes would

result in the TSs being consistent with the current design, as detailed

in the Final Safety Analysis Report, and the current surveillance

procedures.

Specifically, TS Table 3.3-4, Loss of Power, would be changed by

adding a 2.0 [plus or minus] 0.1 second time delay for the 4.16 kV

Emergency Bus UV (UV Relays) level 1--Trip Setpoint and the Allowable

values.

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:

NNECO concludes that these proposed additions to Technical

Specification Table 3.3-4 do not involve a significant hazards

consideration (SHC) and do not involve a significant impact on

public health and safety. The basis for this conclusion is that the

three criteria of 10 CFR 50.92(c) are not compromised. That is, the

proposed changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes will add a time delay, including allowance,

to a portion of the Engineered Safety Feature Actuation System

(ESFAS) Undervoltage (UV) Trip Technical Specification Table 3.3-4.

These changes will align the Technical Specifications to the

existing plant design, as described in the Final Safety Analysis

Report (FSAR) system description and the existing surveillance

procedure. No new plant modifications are associated with this

addition to the Technical Specifications.

The addition of the Level One UV trip time delay setpoint does

not impact any system or component whose failure results in

initiation of the accidents described in the FSAR. Therefore, the

changes do not affect the probability of occurrence of the

previously evaluated accidents. The Level One UV trip time delay

potentially affects the Emergency Diesel Generator (EDG) response

time to accident conditions that occur coincident with a loss of

normal power (LPN). However, previous analysis of the increase in

the time delay (0.5 seconds to 2.0 [plus or minus] 0.1 second)

concluded that the ESFAS response times for those events considered

to occur coincident with an LNP, are not challenged by the time

delay. This conclusion is based upon a comparison between the EDG

start time and the maximum time required to complete those LNP trip

functions necessary to support EDG availability for worst case

accident conditions (Loss of Coolant Accident which results in a

Safety Injection Actuation Signal (SIAS) coincident with LNP). The

calculated EDG start time considered the ESFAS response time (0.5

seconds) in addition to the maximum EDG start time of 15 seconds

after receipt of an SIAS, as specified in Technical Specification

Surveillance Requirement 4.8.1.1.2.a.2. Since the calculated LNP

trip time delay of 15.14 seconds is less than the calculated SIAS

initiated EDG start time of 15.5 seconds, the proposed changes do

not increase the likelihood of an EDG malfunction during an accident

condition. Consequently, the proposed additions do not adversely

affect the ability of either the ESFAS or the EDGs to perform their

intended safety function. The proposed additions to Table 3.3-4 do

not modify the Limiting Condition for Operation or the specific

surveillance procedure acceptance criterion, nor do they change the

frequency of the surveillance. The proposed changes do not involve

any physical changes to the plant and do not alter the way any

structure, system, or component functions. The proposed changes do

not have any adverse impact on the design basis accidents previously

analyzed. The proposed changes do not result in an increase in

radiation exposure to either members of the public or site personnel

because accident mitigation systems will be available consistent

with the assumptions used in the accident analysis. Therefore, the

proposed additions to Technical Specification Table 3.3-4 do not

affect the consequences of the previously evaluated accidents.

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

significant increase in the probability of consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

The function availability and failure modes of equipment

important to safety are

[[Page 2281]]

unaffected by the addition of the 2.0 [plus or minus] 0.1 second

Level One UV trip time delay to Technical Specification Table 3.3-4.

The additions do not introduce any new, credible accidents, or any

new failure modes.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

The proposed additions to Technical Specification Table 3.3-4 do

not have any adverse impact on the accident analyses. Actuation of

the required safety systems is not delayed because the proposed

additions do not delay the time at which the EDGs are required, by

the plant Technical Specifications, to be available to power the

required loads.

Therefore, based on the above, there is no significant reduction

in the margin of safety.

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

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

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

amendment 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 Deputy Director: Phillip F. McKee.

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

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

Jersey

Date of amendment request: December 15, 1997.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TSs) to adopt Option B, of 10 CFR

Part 50, Appendix J, to implement a performance-based approach for Type

B and C testing. Additionally, the wording in the TSs would be modified

for the previous adoption of Option B on Type A testing and a section

added on the primary containment leakage rate testing program.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

Containment leak rate testing is not an initiator of any

accident. The proposed changes do not make any physical changes to

the containment and do not affect reactor operations or the accident

analyses. Therefore, the proposed changes do not involve a

significant increase in the probability of any previously evaluated

accident.

Since the allowable leakage rate is not being changed and since

the analysis documented in NUREG-1493, ``Performance-Based

Containment Leak-Rate Program'' concludes that the impact on public

health and safety due to extended intervals is negligible, the

proposed changes will not involve a significant increase in the

consequences of any previously evaluated accident.

Therefore, adoption of a performance-based leakage testing

requirements will provide an equivalent level of safety and does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

No physical changes are being made to the plant, nor are there

any changes being made to the operation of the plant as a result of

the proposed changes. In addition, no new failure modes of plant

equipment previously evaluated are being introduced.

Therefore, the proposed amendment will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes are based on NRC-approved provisions and

maintain adequate levels of reliability of containment integrity.

The performance-based approach to leakage rate testing recognizes

that historically good results of containment testing provide

appropriate assurance of future containment integrity. This supports

the conclusion that the impact on the health and safety of the

public as a result of extended test intervals is negligible. Since

the analysis documented in NUREG-1493 confirms that the performance

based schedule continues to maintain a minimal impact on public

risk, it can be concluded that the margin of safety is not

significantly affected by the proposed changes.

Therefore, the proposed amendment will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: John F. Stolz.

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: July 23, 1997, as supplemented

September 30 and December 18, 1997. The July 23, 1997, application was

previously noticed in the Federal Register on September 10, 1997 (62 FR

47699). The December 18, 1997, supplement provided additional

information that revised the licensee's evaluation of the significant

hazards consideration. Therefore, renotification of the Commission's

proposed determination of no significant hazards consideration is

necessary.

Description of amendments request: The proposed amendments would

revise the Technical Specifications (TSs) by relocating the reactor

coolant system pressure and temperature limits from the TSs to the

proposed Pressure Temperature Limits Report in accordance with the

guidance provided by Generic Letter 96-03, ``Relocation of the Pressure

Temperature Limit Curves and Low Temperature Overpressure Protection

System Limits.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed removal of the Reactor Coolant System (RCS)

pressure temperature (P-T) limits from the Technical Specifications

(TSs) and relocation to the proposed Pressure Temperature Limits

Report (PTLR) in accordance with the guidance provided by Generic

Letter (GL) 96-03 is administrative in that the requirements for the

P-T limits are unchanged. The P-T limits proposed for inclusion in

the PTLR are based on the fluence associated with 2775 MW thermal

power and operation through 21.9 effective full power years (EFPY)

for Unit 1 and 33.8 EFPY for Unit 2. GL 96-03 requires that the P-T

limits be generated in accordance with the requirements of 10 CFR

[Part] 50, Appendices G and H, documented in an NRC-approved

methodology incorporated by reference in the TSs.

[[Page 2282]]

Accordingly, the proposed curves have been generated using the NRC-

approved methods described in WCAP-14040-NP-A, Revision 2, as

modified at the direction of the NRC Staff, and meet the

requirements of 10 CFR [Part] 50, Appendices G and H. TS 3.4.10.1

will continue to require that the RCS pressure and temperature be

limited in accordance with the limits specified in the PTLR. The

NRC-approval document will be specified in TS 6.9.1.15 and NRC

approval will be required in the form of a TS Amendment prior to

changing the methodology. Use of P-T limit curves generated using

the NRC-approved methods will provide additional protection for the

integrity of the reactor vessel, thereby assuring that the reactor

vessel is capable of providing its function as a radiological

barrier.

TS 3.4.10.3 for Farley Nuclear Plant (FNP) Unit 1 and Unit 2

provides the operability requirements for RCS low temperature

overpressure protection (LTOP). Specifically, TS 3.4.10.3 requires

that two residual heat removal (RHR) system suction relief valves

(RHRRVs) be operable or that the RCS be vented at RCS cold leg

temperatures less than or equal to 310[ deg.]F. Consistent with GL

96-03, the Farley Unit 1 and Unit 2 requirements for LTOP will be

retained in TS 3.4.10.3 and will be evaluated in accordance with the

proposed methodology.

Based on the above evaluation, the proposed changes are

administrative in nature and 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.

As stated above, the proposed changes to remove the RCS P-T

limits from the TSs and relocate them to the proposed PTLR is an

administrative change. Consistent with the guidance provided by GL

96-03, the proposed P-T limits contained in the proposed PTLR meet

the requirements of 10 CFR [Part] 50, Appendices G and H, and were

generated using the NRC-approved methods described in WCAP-14040-NP-

A, Revision 2, as modified at the direction of the NRC Staff. The

proposed changes do not result in a physical change to the plant or

add any new or different operating requirements on plant systems,

structures, or components with the exception of limiting the number

of operating RCPs at RCS temperatures below 110[ deg.]F. Limiting

the number of operating RCPs below 110[ deg.]F results in a

reduction in the [delta]P between the reactor vessel beltline and

the RHRRVs, thereby providing additional margin to limits of

Appendix G. Provisions are made to allow the start of a second RCP

at temperatures below 110[ deg.]F in order to secure the pump that

was originally operating without interrupting RCS flow. The LTOP

enable temperature exceeds the minimum LTOP enable temperature

determined as described in WCAP-14040-NP-A, Rev. 2, thereby

providing additional assurance that the LTOP system will be

available to protect the RCS in the event of an overpressure

transient at RCS temperatures at or below 310[ deg.]F. Based on the

methods contained in WCAP-14040-NP-A, Rev. 2, the minimum boltup

temperature for the reactor vessel flange region is conservatively

established as 70[ deg.]F.

As stated in the above response, implementation of the proposed

changes do not result in a significant increase in the probability

of a new or different accident (i.e., loss of reactor vessel

integrity). The RCS P-T limits will continue to meet the

requirements of 10 CFR [Part] 50, Appendices G and H, and will be

generated in accordance with the NRC approved methodology described

in WCAP-14040-NP-A, Revision 2, as modified at the direction of the

NRC Staff. Therefore, the proposed changes do not result in a

significant increase in the possibility of a new or different

accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The margin of safety is not affected by the removal of the RCS

P-T limits from the TSs and relocating them to the proposed PTLR.

The RCS P-T limits will continue to meet the requirements of 10 CFR

50, Appendices G and H. To provide additional assurance that the P-T

limits continue to meet the requirements of Appendices G and H, TS

6.9.1.15 will require the use of the NRC-approved methodology to

generate P-T limits. The RCS LTOP requirements will be retained in

TS 3.4.10.3 due to use of the RHRRVs for LTOP, consistent with the

guidance provided by GL 96-03, and will be verified to provide

adequate protection of the reactor coolant system against the limits

of Appendix G. The LTOP enable temperature exceeds the LTOP enable

temperature determined in accordance with the NRC-approved

methodology, thus protecting the RCS in the event of a low

temperature overpressure transient over a broader range of

temperatures than required by WCAP-14040-NP-A, Rev. 2.

Administrative procedures will preclude operation of the RCS at

temperatures below the minimum boltup temperature for the reactor

vessel head, thus precluding the possibility of tensioning the

reactor vessel head at RCS temperatures below the minimum boltup

temperature. Operation of the plant in accordance with the RCS P-T

limits specified in the PTLR and continued operation of the LTOP

system in accordance with TS 3.4.10.3 will continue to meet the

requirements of 10 CFR [Part] 50, Appendices G and H, and will

therefore, assure that a margin of safety is not significantly

decreased as the result of the proposed changes.

Based on the preceding analysis, SNC [Southern Nuclear Operating

Company] has determined that removal of the RCS P-T limits from the

TS and relocation to the proposed PTLR will not significantly

increase the probability or consequences of an accident previously

evaluated, create the possibility of a new or different kind of

accident from any accident previously evaluated, or involve a

significant reduction in a margin of safety. SNC therefore concludes

that the proposed change meets the requirements of 10 CFR 50.92(c)

and 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: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama.

NRC Project Director: Herbert N. Berkow.

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

Callaway County, Missouri

Date of application request: October 17, 1997.

Description of amendment request: The proposed amendment would

modify the technical specifications (TS) for plant heatup and cooldown

curves and the maximum allowable power operated relief valve (PORV)

setpoint curve for cold overpressure protection, as found in TS Figures

3.4-2, 3.4-3, and 3.4-4. These changes are requested to incorporate

information gained from Surveillance Capsule V, which was removed

during Callaway Refuel 8 in the fall of 1996 after 9.85 effective full

power years (EFPY) of exposure. Capsule V is the third capsule to be

removed from the reactor vessel in the continuing surveillance program

that monitors the effects of neutron irradiation on the Callway reactor

vessel materials under actual plant operating conditions. The proposed

changes include:

(1) Figure 3.4-2, heatup limitation curve and Figure 3.4-3,

cooldown limitation curve, would be revised to reflect the

TRNDT calculated for 20 EFPY in the surveillance capsule

report.

(2) Figure 3.4-4 is the maximum allowable PORV setpoint curve for

cold overpressure protection. This curve would be (a) revised to

account for the changes made in the heatup and cooldown limitation

curves, (b) allow for the operation of the normal charging pump, and

(c) account for instrument accuracy and other uncertainties.

(3) TS Bases 3/4.4.9 and 3/4.5.2 through 3/4.5.4 would be revised

by correcting miscellaneous items and by adding discussion of the

normal charging pump.

Basis for proposed no significant hazards consideration

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

provided its analysis of the

[[Page 2283]]

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.

Pressure and temperature limits for the reactor pressure vessel

(RPV) are established to the requirements of 10 CFR 50, Appendix G

to ensure brittle fracture of the vessel does not occur. This

amendment revises the P/T curves in the TS to reflect the material

capsule surveillance results from the sample removed during the fall

outage of 1996.

The RPV surveillance capsule contained flux wires for neutron

flux monitoring and Charpy V notch impact and tensile test

specimens. The irradiated material properties were compared to

available unirradiated properties to determine the effect of

irradiation on material toughness for the base and weld materials

through Charpy testing. Irradiated tensile testing results are

compared with unirradiated data to determine the effect of

irradiation on the stress-strain relationship of the materials.

The P/T curves are modified to reflect the results of the above

examination. These curves and their operating limits were generated

using the NRC-approved methods described in WCAP-14040-NP-A,

Revision 2 and meet the requirements of 10 CFR 50, Appendices G and

H as modified by the provisions of ASME Code Case N-514. The new

curves therefore represent the latest information available on the

state of the reactor vessel materials. The P/T curves are generated

for reactor vessel protection against brittle fracture, they do not

affect the recirculation piping. Accordingly, the probability of

occurrence of a design basis Loss of Coolant Accident (LOCA) is not

increased. Likewise, no other previously evaluated accident and

transients, as defined in Chapter 15 of the Final Safety Analysis

Report are affected by this proposed change to the Callaway P/T

curves. Additionally, this proposed revision does not affect the

design, operation, or maintenance of any safety-related system

designed for the mitigation or prevention of previously analyzed

events.

Since no previously evaluated accidents or transients are

affected by this change, their probability of occurrence and

consequences is not increased.

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

or different kind of accident from any accident previously

evaluated.

Implementing the proposed P/T curves into the TS does not alter

the design or operation of any system or piece of equipment designed

for the prevention or mitigation of accidents and transients. As a

result, no new operating modes are introduced from which a new type

accident becomes possible. Existing systems will continue to be

operated per present design basis assumptions.

The proposed P/T limits were generated from the evaluation of

the material capsule removed during the fall outage of 1996 using

the NRC-approved methods described in WCAP-14040-NP-A, Revision 2.

As a result, these limits include the latest available information

on the reactor vessel materials. Furthermore, they will continue to

be monitored per the requirements of the TS and 10 CFR 50,

Appendices G and H. For the above reasons, the changes do not create

the possibility of a new type of accident.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The purpose of the P/T limits is to avoid a brittle fracture of

the reactor vessel. As such, material capsules are removed

periodically to determine the effects of neutron irradiation on the

reactor vessel materials. This change to the Callaway curves is

proposed to incorporate the evaluation results of the latest capsule

removed during the fall outage of 1996. Accordingly, these curves

represent the latest information available on the reactor vessel

materials.

Also, the curves were generated using the approved methodologies

of 10 CFR 50, Appendix G.

The Cold Overpressure Mitigation System Curve (Figure 3.4-4) is

also revised to reflect exposure dependencies. This curve was

generated for 20 EFPY using approved methodologies and reflects the

results of this latest material capsule report. Utilizing the

methodology set forth in ASME Section XI, Appendix G, which includes

the provisions of Code Case N-514, and 10 CFR 50, Appendices G and H

ensures that proper limits and conservative safety factors are

maintained.

The proposed changes do not affect the evaluation of any FSAR

Chapter 15 transient and accident. Furthermore, the proposed change

does not affect the operation of systems or equipment important to

safety.

The Limiting Condition for Operation of Specification 3.4.9 will

not change. Also, no TS surveillance or surveillance frequencies are

revised as a result of this Technical Specification submittal,

besides the fact that the P/T surveillance will now refer to the

revised curves. Procedures regarding the monitoring of the P/T

limits during reactor startup, cooldown, and leakage testing will

not change as a result of this proposed Technical Specification

change with respect to frequency of the surveillance or the methods

used to perform the surveillance. Thus, the P/T limits will continue

to be surveilled as before per the same procedures and at the same

frequencies.

No other Technical Specifications are affected by the proposed

revision. The margin of safety to any Technical Specification safety

limit therefore is not reduced. For the above reasons the new curves

do not represent a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

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

Callaway County, Missouri

Date of application request: October 31, 1997.

Description of amendment request: The proposed amendment would

revise the engineered safety features actuation system (ESFAS)

Functional Unit 6.f, Loss of Offsite Power-Start Turbine-Driven Pump,

in Technical Specification Tables 3.3-3, 3.3-4, and 4.3-2. The tables

would be revised to create separate functional units for the analog and

digital portions of the ESFAS function associated with starting the

turbine-driven auxiliary feedwater pump (TDAFP) upon a loss of offsite

power.

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.

Overall protection system performance will remain within the

bounds of the previously performed accident analyses since no

hardware changes are proposed. The recognition that different

operability and surveillance requirements apply to analog vs.

digital circuitry does not impact any previously analyzed accidents.

The proposed change will not affect any of the analysis assumptions

for any of the accidents previously evaluated. The proposed change

does not alter the current method or procedures for meeting the

surveillance requirements in Table 4.3-2. The proposed change will

not affect the probability of any event initiators nor will the

proposed change affect the ability of any safety-related equipment

to perform its intended function. There will be no degradation in

the performance of nor an increase in the number of challenges

imposed on safety-related equipment assumed to function during an

accident situation. Therefore, the proposed change does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

There are no hardware changes nor are there any changes in the

method by which any safety-related plant system performs its

[[Page 2284]]

safety function. The separation of analog and digital portions of

Functional Unit 6.f will not impact the normal method of plant

operation.

The operability requirements, ACTION Statement, and surveillance

requirements for the analog portion, new Functional Unit 6.f.1), are

identical to those of Functional Unit 8.a, while the requirements

for the digital portion, new Functional Unit 6.f.2), are consistent

with the current Technical Specifications, other than the new ACTION

Statement 39 provisions that defer to the TDAFP Specification

3.7.1.2 requirements and the performance of a TADOT during

appropriate plant conditions. These changes do not change any ESFAS

design standards and are appropriate for digital functions such as

this. No new accident scenarios, transient precursors, failure

mechanisms, or limiting single failures are introduced as a result

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

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change does not affect the acceptance criteria for

any analyzed event. There will be no effect on the manner in which

safety limits or limiting safety system settings are determined nor

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

the accomplishment of protection functions. There will be no impact

on any margin of safety.

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

this review, it appears that the three standards of 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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

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

Callaway County, Missouri

Date of application request: October 31, 1997.

Description of amendment request: The proposed amendment would

change Technical Specification Tables 2.2-1, 4.3-1, and 3.3-4, as well

as their associated Bases, in order to reduce repeated alarms, rod

blocks, and partial reactor trips that continue to manifest themselves,

especially during beginning of cycle operation following refueling

outages. Besides the potential for distracting operator attention away

from more safety significant evolutions, these occurrences have also

led to power reductions during surveillance testing in order to avoid

reactor trips, since the channel being tested is placed in the tripped

condition. These changes to various setpoint terms associated with the

overtemperature delta T, overpower delta T, and steam generator (SG)

water level low-low vessel delta T (Power-1 and Power-2) reactor trip

and auxiliary feedwater (AFW) start engineered safety feature actuation

system (ESFAS) functions will improve plant operations and reduce the

potential for unnecessary reactor trips, with no detrimental effect on

the plant's safety analysis or licensing basis.

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.

Overall protection system performance will remain within the

bounds of the previously performed accident analyses since no

hardware changes are proposed. The protection systems will continue

to function in a manner consistent with the plant design basis. The

proposed changes will not affect any of the analysis assumptions for

any of the accidents previously evaluated. The proposed changes will

not affect the probability of any event initiators nor will the

proposed changes affect the ability of any safety-related equipment

to perform its intended function. There will be no degradation in

the performance of nor an increase in the number of challenges

imposed on safety-related equipment assumed to function during an

accident situation. There will be no change to normal plant

operating parameters or accident mitigation capabilities. Therefore,

the 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. There are no hardware changes associated with this

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

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

normal manner of plant operation is unchanged. The Overtemperature

delta T Allowable Value increase is justified by the use of existing

setpoint margin and elimination of conservatisms not required by the

safety analysis and licensing basis. There will be a reduction in

the incidence of alarms, rod stops, and partial reactor trips. There

will also be less of a need to reduce power during on-line

surveillance testing. These changes represent substantial plant

operational improvements.

No new accident scenarios, transient precursors, failure

mechanisms, or limiting single failures are introduced as a result

of these changes. There will be no adverse effect or challenges

imposed on any safety-related system as a result of these changes.

Therefore, the 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 a margin of safety.

The proposed changes do not affect the acceptance criteria for

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

Limit (SAL). Maintaining the SAL preserves the margin of safety.

There will be no effect on the manner in which safety limits or

limiting safety system settings are determined nor will there be any

effect on those plant systems necessary to assure the accomplishment

of protection functions. There will be no impact on the overpower

limit, DNBR limits, FQ, F(delta)H, LOCA PCT, peak local

power density, or any other margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of amendment request: December 4, 1997.

Description of amendment request: This amendment as reflected in

Section 2.1.1.2 of the Technical Specifications would continue the use

of the existing Siemens Power Corporation minimum critical power ratio

(MCPR) safety limits for Cycle 14 and would change the Asea Brown

Boveri (ABB) MCPR safety limit for single loop operation from 1.08 for

Cycle 13 to 1.09 for Cycle 14.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or

[[Page 2285]]

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. The proposed

Technical Specifications amendment continues the use of

conservatively established ATRIUM-9X MCPR safety limits for WNP-2

such that the fuel is protected during normal operation as well as

during plant transients or anticipated operational occurrences.

The probability of an evaluated accident is not increased by the

continued use of the interim ATRIUM-9X MCPR safety limit of 1.13

(two loop operation) or 1.14 (single loop operation) or from

changing the ABB single loop MCPR safety limit of 1.08 (Cycle 13) to

1.09 (Cycle 14). The changes do not require any physical plant

modifications, physically affect any plant component, or entail

changes in plant operation. The increase in single loop MCPR safety

limit is attributed to a slightly more conservative assembly power

distribution used in the Cycle 14 calculations following ABB

standard methodology. While the Cycle 13 result is also

conservative, the increase in Cycle 14 is intended to accommodate

small cycle to cycle variability. Therefore, no individual

precursors of an accident are affected.

This Technical Specification amendment proposes to continue

using the interim MCPR safety limits for ATRIUM-9X fuel to protect

the fuel during normal operation as well as during plant transients

or anticipated operational occurrences. The method that is used to

determine the ATRIUM-9X additive constant uncertainty is

conservative, such that the resulting interim ATRIUM-9X MCPR safety

limits are high enough to ensure that less than 0.1% of the fuel

rods are expected to experience boiling transition if the limit is

not violated. Using NRC approved methodology, ABB has utilized these

interim values as the basis for the Cycle 14 safety limit for the

co-resident ATRIUM-9X. Operational limits have been established

based on the interim ATRIUM-9X MCPR safety limits to ensure that the

safety limits are not violated. This will ensure that the fuel

design safety criteria (more than 99.9% of the fuel rods avoid

transition boiling during normal operation as well as anticipated

operational occurrences) is met. In addition, since the operability

of plant systems designed to mitigate any consequences of accidents

have not changed, the consequences of an accident previously

evaluated are not expected to increase.

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

different kind of accident from any accident previously evaluated?

Creation of the possibility of a new or different kind of

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

that accident. New accident precursors may be created by

modifications of the plant configuration, including changes in

allowable modes of operation. This Technical Specification submittal

does not involve any modifications of the plant configuration or

allowable modes of operation. This Technical Specification change

continues the use of added conservatism in the ATRIUM-9X MCPR safety

limits which resulted from analytical changes and use of an expended

database. Also, ABB has calculated single loop MCPR safety limit

[which is] about 0.006 greater in Cycle 14 than was used in Cycle

13. The increase in single loop MCPR safety limit is attributed to a

slightly more conservative assembly power distribution used in the

Cycle 14 calculations following ABB standard methodology. While the

Cycle 13 result is also conservative, the increase in Cycle 14 is

intended to accommodate small cycle to cycle variability. Therefore,

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

kinds of accidents are created.

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

margin of safety?

The continued use of interim MCPR safety limits provides a

margin of safety by ensuring that less than 0.1% of the rods are

expected to be in boiling transition if the MCPR limit is not

violated. These interim limits are based on calculations by SPC

using the revised ATRIUM-9X additive constant uncertainty. These

calculations are based on a larger pool of data than previous

calculations (527 data points versus 82 data points). Additionally,

the revised additive constant uncertainty has been conservatively

applied in the calculation of the interim ATRIUM-9X MCPR safety

limits resulting in more restrictive limits.

The calculated single loop MCPR safety limit results are about

0.006 greater for Cycle 14 than they were for Cycle 13. The increase

in single loop MCPR safety limits is attributed to a slightly more

conservative assembly power distribution used in the Cycle 14

calculations following ABB standard methodology. Because the fuel

design safety criteria of more than 99.9% of the fuel rods avoiding

transition boiling during normal operation as well as anticipated

operational occurrences is met, there is not a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005-3502.

NRC Project Director: William H. Bateman.

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.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear

Generating Station, Unit No. 1, Salem County, New Jersey

Date of amendment request: December 11, 1997.

Brief description of amendment request: The proposed amendment

would provide a one-time change to the Technical Specifications to

allow purging of the containment during Modes 3 (Hot Standby) and 4

(Hot Shutdown) upon return to power from the current refueling outage

(1R13).

Date of publication of individual notice in Federal Register:

December 18, 1997 (62 FR 66397).

Expiration date of individual notice: January 20, 1998.

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

West Broadway, Salem, NJ 08079.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination,

[[Page 2286]]

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.

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

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

Date of application for amendments: May 6, 1997, as supplemented on

July 30, 1997.

Brief description of amendments: The amendments will change

Technical Specification 3/4.7.5, ``Ultimate Heat Sink'' and the

associated Bases to support steam generator replacement and incorporate

recent Ultimate Heat Sink (UHS) design evaluations.

Date of issuance: December 12, 1997.

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

Amendment Nos.: 95 and 95.

Facility Operating License Nos. NPF-37 and NPF-66: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 2, 1997 (62 FR

35847). The July 30, 1997, submittal provided additional clarifying

information that did not change the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 12, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Byron Public Library District,

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

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

Date of application for amendments: September 30, 1997.

Brief description of amendments: The amendments would add a new

Technical Specification (TS) Section 3/4.12.C and associated bases to

allow certain reactor coolant pressure tests to be performed in MODE 4

when the reactor pressure vessel requires testing at temperatures

greater than 212 degrees Fahrenheit. This temperature normally

corresponds with MODE 3.

Date of issuance: January 5, 1998.

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

Amendment Nos.: 164, 159, 179 and 177.

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: November 19, 1997 (62

FR 61839). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 5, 1998.

No significant hazards consideration comments received: No.

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.

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

No. 2, Pope County, Arkansas

Date of application for amendment: April 24, 1997.

Brief description of amendment: The amendment revises the inservice

inspection requirements associated with steam generator tube sleeves.

Date of issuance: December 23, 1997.

Effective date: December 23, 1997.

Amendment No.: 187.

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

Technical Specifications.

Date of initial notice in Federal Register: July 16, 1997 (62 FR

38134).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 23, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana

Date of amendment request: October 15, 1997.

Brief description of amendment: The amendment revises the license

to reflect the transfer of the 30-percent undivided ownership interest

in the River Bend Station, Unit No. 1 from Cajun Electric Power

Cooperative, Inc. to Entergy Gulf States, Inc. The transfer was

approved by Order dated November 28, 1997, which was published in the

Federal Register on December 5, 1997 (62 FR 64404).

Date of issuance: December 23, 1997.

Effective date: December 23, 1997.

Amendment No.: 101.

Facility Operating License No. NPF-47: The amendment revised the

operating license.

Date of initial notice in Federal Register: October 24, 1997 (62 FR

55432).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 23, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Documents

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

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: July 18, 1997, as supplemented

September 12 and October 25, 1997.

Brief description of amendment: Establish a new Low-Temperature

Overpressure Protection Technical Specification (TS).

Date of issuance: December 22, 1997.

Effective date: December 22, 1997.

Amendment No.: 161.

Facility Operating License No. DPR-31: Amendment revised the TSs.

Date of initial notice in Federal Register: August 13, 1997 (62 FR

43369) The supplemental letters dated September 12 and October 25, 1997

did not change the initial no significant hazards consideration

determination or expand the scope of the initial notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1997.

[[Page 2287]]

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal River, Florida 34428.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: July 29, 1997, as supplemented

October 29, 1997.

Brief description of amendment: The amendment revises the Post-

Accident Monitoring Instrumentation Technical Specification (TS).

Date of issuance: December 22, 1997.

Effective date: December 22, 1997.

Amendment No.: 162.

Facility Operating License No. DPR-31: Amendment revised the TSs.

Date of initial notice in Federal Register: August 13, 1997 (62 FR

43369).

The supplemental letter October 29, 1997 did not change the initial

no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal River, Florida 34428.

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

Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: February 24, 1997, as

supplemented on April 24 and December 4, 1997.

Brief description of amendments: The admendments change technical

specification section 6.9.1.7, Core Operating Limits Report, to reflect

use of the Westinghouse Best Estimate Large Break Loss-of-Coolant

Accident (LOCA) methodology for large break LOCA analysis, including

supporting documents.

Date of issuance: December 20, 1997.

Effective date: December 20, 1997.

Amendment Nos.: 195 and 189.

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

revised the TS.

Date of initial notice in Federal Register: June 4, 1997 (62 FR

30631). By letter dated December 4, 1997, the licensee provided

additional information which did not affect the original no significant

hazards determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 20, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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: October 7, 1997, as supplemented

December 17, 1997.

Brief description of amendment: Technical Specifications 4.6.1.1,

3/4.6.1.2, and 3/4.6.1.3 require the testing of the containment to

verify leakage limits at a specified test pressure. The amendment (1)

modifies the list of valves that can be opened in Modes 1 through 4,

(2) removes a footnote on Type A testing, and (3) makes editorial

changes to the Technical Specifications and makes changes to the

associated Bases sections.

Date of issuance: December 18, 1997.

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

within 60 days.

Amendment No.: 154.

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

Technical Specifications. Date of initial notice in Federal Register:

November 5, 1997 (62 FR 59917) The December 17, 1997, letter

provide clarifying information that did not change the October 7, 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 December 18, 1997.

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 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear

Generating Station, Unit No. 1, Salem County, New Jersey

Date of application for amendment: October 6, 1997.

Brief description of amendment: The requested changes would

increase the allowable band for control and shutdown rod demanded

position versus indicated position from plus or minus 12 steps to plus

or minus 18 steps when the power level is not greater than 85% rated

thermal power. The changes have already been approved for Salem Unit 2

in Amendment No. 183, issued September 10, 1997, as an exigent

amendment.

Date of issuance: December 22, 1997.

Effective date: December 22, 1997.

Amendment No.: 202.

Facility Operating License No. DPR-70: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 19, 1997 (62

FR 61845).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1997.

No significant hazards consideration comments received: No.

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

West Broadway, Salem, NJ 08079.

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

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

Date of application for amendment: June 19, 1997, with additional

information provided on August 15, 1997 (TS 391T).

Brief Description of amendment: The amendments revise the Technical

Specifications (TS) to temporarily extend the allowed outage time for

the emergency diesel generators from 7 to 14 days to permit completion

of preventive maintenance.

Date of issuance: December 22, 1997.

Effective Date: December 22, 1997.

Amendment Nos.: 250 and 209.

Facility Operating License Nos. DPR-52 and DPR-68: Amendment

revised the TS.

Date of initial notice in Federal Register: July 30, 1997 (62 FR

40858).

The additional information provided on August 15, 1997 does not

affect the staff's proposed finding of no significant hazards

consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1997.

No significant hazards consideration comments received: None.

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

South Street, Athens, Alabama 35611.

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

Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: July 9, 1996 (TXXX-96393), as

supplemented

[[Page 2288]]

on December 12, 1997 (TXXX-97268). (The supplement contains clarifying

information and does not change the staff's original proposed no

significant hazards determination.)

Brief description of amendments: The amendments change Technical

Specification 3.3-3, ``Engineered Safety Features Actuation System

Instrumentation Trip Setpoints.'' The proposed changes would increase

the minimum allowable value of the Unit 1 Steam Line Pressure--Low

Safety Injection and Steam Line Isolation functions. These changes are

needed to ensure that the instrumentation error is properly accounted

for in the TSs.

Date of issuance: December 30, 1997.

Effective date: December 30, 1997, to be implemented within 30

days.

Amendment Nos.: Unit 1--Amendment No. 56; Unit 2--Amendment No. 42.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

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

FR 6579) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 30, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

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

19497, Arlington, TX 76019.

Dated at Rockville, Maryland, this 7th day of January 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects--III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 98-753 Filed 1-13-98; 8:45 am]

BILLING CODE 7590-13-P