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

[[Page 48257]]

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 August 17, 1998, through August 28, 1998.

The last biweekly notice was published on August 26, 1998 (63 FR

45521).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Administration Services, Office of

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

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

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

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

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

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By October 9, 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

[[Page 48258]]

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.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: August 17, 1998.

Description of amendment request: The Carolina Power & Light

Company, licensee for the Brunswick Steam Electric Plant (BSEP), Unit

Nos. 1 and 2, proposed amendments to the Technical Specifications (TS)

to revise the requirement that the operations manager hold or has held

a senior reactor operator (SRO) license. The proposed revision would

require that either the operations manager or assistant operations

manager hold an SRO license.

The licensee has concluded that the proposed license amendments do

not involve a Significant Hazards Consideration. In support of this

determination, an evaluation of each of the three standards set forth

in 10 CFR 50.92 is provided below.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would operation of the facility in accordance with the

proposed amendment involve a significant increase in the probability

or consequences of an accident previously evaluated?

The proposed license amendments do not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The change to Technical Specification 5.2.2.f

to require the operations manager or assistant operations manager to

hold an SRO license is administrative in nature and does not

directly affect plant operations. The change does not physically

alter the facility in any manner and, as such, does not affect the

means in which any safety-related system performs its intended

safety function.

2. Would operation of the facility in accordance with the

proposed amendment create the possibility of a new or different kind

of accident from any accident previously evaluated?

The proposed license amendments will not create the possibility

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

evaluated. As stated above, the proposed change is administrative in

nature. It does not involve physical alterations of the plant

configuration or changes in setpoints or operating parameters.

Therefore, there is no possibility of creating a new or different

kind of accident.

3. Would operation of the facility in accordance with the

proposed amendment involve a significant reduction in a margin of

safety?

The proposed license amendments do not involve a significant

reduction in a margin of safety. The proposed change to Technical

Specification 5.2.2.f, requiring the operations manager or assistant

operations manager to hold an SRO license is consistent with (1) 10

CFR 50.54(l), which requires individuals responsible for directing

the licensed activities of licensed operators to hold an SRO

license, (2) the previously approved wording of Revision 1 of NUREG-

1433, ``Standard Technical Specifications General Electric Plants,

BWR/4,'' and Technical Specification Traveler Form (TSTF) 65,

Revision 1, and (3) the intent of ANSI-N18.1.-1971, ``Selection and

Training of Nuclear Power Plant Personnel.'' Therefore, the proposed

change does not represent a reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Pao-Tsin Kuo (Acting).

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 14, 1998.

Description of amendment request: The proposed amendments would

change the Dresden, Quad Cities, and LaSalle Technical Specifications

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

fuel. Specifically the proposed amendments incorporate the following

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

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

administrative changes that eliminate the cycle specific implementation

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

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

Power Ratio (MCPR). This amendment request supersedes in its entirety a

letter from J. Hosmer (ComEd) to U.S. NRC, ``Technical Specification

Changes for Transition to Siemens Power Corporation ATRIUM-9B Fuel,''

dated August 29, 1997 (63 FR 2274).

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.

[[Page 48259]]

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

The Reference 1 methodology to be added to the Technical

Specifications is used as part of the LOCA 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 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 limits. The Technical Specification requirements

for monitoring APLHGR are not affected by this change. The revised

method will result in higher APLHGR limits, thus the SPC fuel will

be allowed to operate at higher nodal powers. The approved

methodology, however, still protects the fuel performance limits

specified by 10 CFR 50.46. Therefore, the probability or

consequences of an accident previously evaluated will not change.

b. Addition of SPC Generic Methodology for Application of ANFB 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 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 fuel. Therefore, it provides data that is used

in the determination of the MCPR Safety Limit. This approved

methodology for applying the ANFB critical power correlation to the

GE fuel will protect the fuel from boiling transition. Operational

MCPR limits will also be applied to ensure that the MCPR Safety

Limit is protected during all modes of operation and anticipated

operational occurrences. Because Reference 3 contains conservative

methods and calculations and because the operability of plant

systems designed to mitigate any consequences of accidents have not

changed, the probability or consequences of an accident previously

evaluated will not increase.

c. Addition of SPC Topical for Revised ANFB Correlation Uncertainty

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

and 2)

The probability or consequences of a previously evaluated

accident are not increased by adding Reference 7 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. Approval of Reference 7 (Reference 20)

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

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

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

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

boiling during normal operation as well as anticipated operational

occurrences. This change does not require any physical plant

modifications, physically affect any plant components, or entail

changes in plant operation. This methodology for determining the

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

calculation will continue to support protecting the fuel from

boiling transition. Operational MCPR limits will be applied to

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

operation and anticipated operational occurrences. Therefore, no

individual precursors of an accident are affected and the

operability of plant systems designed to mitigate the probability or

the consequences of an accident previously evaluated is not affected

by these changes.

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

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

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

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

probability or the consequences of an accident previously evaluated.

This change implements the MCPR Safety Limits resulting from the SPC

ANFB critical power correlation methodology using the ATRIUM-9B

additive constant uncertainty resulting from approval of Reference 7

(Reference 20). The MCPR Safety Limits for Quad Cities Units 1 and

2, Dresden Unit 3, and LaSalle Units 1 and 2 are anticipated to be

conservative and acceptable for future cycles. Cycle specific MCPR

Safety Limit calculations will be performed, consistent with SPC's

approved methodology, to confirm the appropriateness of the MCPR

Safety Limit. Additionally, operational MCPR limits will be applied

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

modes of operation and anticipated operational occurrences. The MCPR

Safety Limits are being set at the CPR value where less than 0.1% of

the rods in the core are expected to experience boiling transition.

These Safety Limits are expected to be applicable for future cycles

of ATRIUM-9B. Therefore the probability or consequences of an

accident will not increase.

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

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

The removal of footnotes from the Quad Cities and Dresden

Technical Specifications does not involve any significant increase

in the probability or consequences of an accident previously

evaluated. The footnotes were added to clarify that cycle specific

methods were used until the generic methodology was approved by the

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

application of the ANFB correlation to the coresident GE fuel

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

database used to calculate the ATRIUM-9B additive constant

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

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

the Quad Cities Technical Specifications is justified by the removal

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

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

physically affect any plant components or entail changes in plant

operation. Therefore, the probability or consequences of an accident

previously evaluated are not expected to increase.

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

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

The revision to the Section 3 Technical Specification

description of the APLHGR limits has no implications on accident

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

allow flexibility for the MAPLHGR limits and their exposure basis to

be specified in the COLR and to establish consistency with approved

methodologies currently utilized by Siemens Power Corporation, which

calculate MAPLHGR limits based on bundle or planar average

exposures. This revision also provides for consistency in the APLHGR

limit Technical Specification wording between the ComEd BWRs. The

revision to the 3.11.D SLHGR 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

limits and their exposure basis to be specified in the COLR. This

revision makes the Dresden LHGR definition consistent with NUREG

1433/1434, Revision 1 wording. The definition of the Average Planar

Exposure is deleted, because the exposure basis of the APLHGR and

LHGR is being removed. Therefore, no plant equipment or processes

are affected by this change. Thus, there is no alteration in the

probability or consequences of an accident previously evaluated.

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

from any accident previously evaluated:

Creation of the possibility of a new or different kind of

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

that accident. New accident precursors may be created by

modifications to the plant configuration, including changes in

allowable modes of operation. This Technical Specification submittal

does not involve any modifications to the plant configuration or

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

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

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

new or different kind of accident from any accident previously

evaluated.

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

The revised jet pump model methodology will be used to analyze

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

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

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

event and determine the MAPLHGR limits. Therefore, the possibility

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

[[Page 48260]]

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

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

LaSalle Units 1 and 2)

Addition of the generic methodology for the application of the

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

the Quad Cities Technical Specifications and Bases Section 2.1.2 and

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

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

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

involves adding an NRC approved methodology, which is used to

determine the additive constants and additive constant uncertainty

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

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

or different kinds of accidents are created.

c. Addition of SPC Topical for Revised ANFB Correlation Uncertainty

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

and 2)

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

the Quad Cities and Dresden Technical Specifications and Bases

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

Specifications will not create the possibility of a new or different

kind of accident from any accident previously evaluated. This

methodology describes the calculation of an input to the MCPR Safety

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

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

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

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

kinds of accidents are created.

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

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

Changing the MCPR Safety Limit will not create the possibility

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

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

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

of the rods avoid boiling transition.

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

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

the ATRIUM-9B additive constant uncertainty resulting from approval

of Reference 7 (Reference 20). The new MCPR Safety Limit for Quad

Cities Units 1 and 2, Dresden Unit 3, and LaSalle Units 1 and 2 are

greater than the current values at Quad Cities Units 1 and 2,

Dresden Unit 3, and LaSalle Units 1 and 2 and are being increased

now in anticipation of bounding future reloads of ATRIUM-9B. This

change does not introduce any physical changes to the plant, the

processes used to operate the plant, or allowable modes of

operation. Therefore, no new accidents are created that are

different from any accident previously evaluated.

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

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

The removal of the footnotes from the Quad Cities and Dresden

Technical Specifications does not create a new or different kind of

accident from any accident previously evaluated. The removal of the

footnotes does not affect plant systems or operation. The footnotes

were temporarily established to implement a conservative cycle

specific MCPR Safety Limit until the SPC generic methodology was

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

are no longer applicable. Removing these footnotes does not

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

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

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

Technical Specifications, which is justified by the removal of the

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

from any accident previously evaluated.

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

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

The revision of the APLHGR and LHGR limit descriptions will not

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

any accident previously evaluated. This revision will not alter any

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

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

be specified in the COLR and to maintain consistency with the

calculated results of methodologies currently used to determine the

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

because it is being removed from LHGR and APLHGR Technical

Specifications. This change does not introduce any physical changes

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

modes of operation. Therefore this change does not create the

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

previously evaluated.

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

the following reasons:

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

The revised jet pump model methodology, and the MAPLHGRs,

resulting from the revised jet pump methodology, will continue to

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

of LOCA analyses performed with this methodology must continue to

comply with the requirements of 10 CFR 50.46. Therefore, there is no

significant reduction in the margin of safety.

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

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

LaSalle Units 1 and 2)

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

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

Bases Section 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 topicals for coresident fuel (Reference 4 and 5). The

MCPR Safety Limit will continue to ensure 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.

c. Addition of SPC Topical for Revised ANFB Correlation Uncertainty

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

and 2)

The MCPR Safety Limit provides a margin of safety by ensuring

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

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

Specification amendment request proposes to insert the topical

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

constant uncertainty. The new ATRIUM-9B additive constant

uncertainty calculation is conservative and is based on a larger

database than previous calculations. Because the criteria of

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

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

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

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

to the Technical Specifications. In addition, operational limits

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

all modes of operation. This revised methodology will ensure that

the appropriate level of fuel protection is being employed.

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

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

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

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

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

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

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

not violated. The proposed Technical Specification amendment request

reflects the MCPR Safety Limit results from conservative evaluations

by SPC using the ANFB critical power correlation with the ATRIUM-9B

additive constant uncertainty resulting from approval of Reference 7

(Reference 20).

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.

[[Page 48261]]

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

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

The removal of the cycle specific footnotes in Quad Cities and

Dresden Technical Specifications does not impose a change in the

margin of safety. These footnotes were added due to concerns

regarding the calculation of the additive constant uncertainty for

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

critical power correlation to coresident GE fuel in Quad Cities Unit

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

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

the topical report describing the increased database used to

calculate the additive constant uncertainties for ATRIUM-9B

(Reference 7) has also received NRC approval (Reference 20) and both

are proposed to be added to the Technical Specifications in this

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

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

Quad Cities Technical Specifications is justified by the removal of

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

B2-3a, also does not impose a change in the margin of safety.

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

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

The revision to the APLHGR and LHGR limit descriptions will not

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

calculate the APLHGR must comply with the guidelines of Appendix K

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

be maintained within the limits specified in the COLR. The

surveillance requirements for these two thermal limits remain

unchanged. Thus, there will be no reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: For Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, IL 60450; for Quad

Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, IL 61021; and

for LaSalle, the Jacobs Memorial Library, 815 North Orlando Smith

Avenue, Illinois Valley Community College, Oglesby, IL 61348-9692.

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

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

NRC Project Director: Stuart A. Richards.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 17, 1996.

Description of amendment request: The proposed change extends the

surveillance interval for the Reactor Trip Breakers (RTBs) from monthly

to quarterly and increases the allowed outage time for operation with

an inoperable RTB from one hour to two hours.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of any accident previously evaluated?

Response: No.

The proposed change to increase RTB surveillance interval will

have no significant effect on the probability or consequences of any

accident previously evaluated. As previously stated, all of the

transient and accident analyses that call for a reactor trip assume

that the reactor trip breakers (RTBs) operate and interrupt power to

the control element drive mechanism (CEDMs). Extensive testing

results, indicate that the RTBs are available and capable of

performing their safety-related function. Currently RTBs are

verified operable every 4 weeks. Under the proposed change RTBs

would be verified operable at least every 6 weeks. This reduced

testing frequency is intended to increase component reliability. The

increase in the testing interval cannot increase component failure

rate or the potential for component failure.

The proposed change to increase the allowed outage time for RTBs

from 1 hour to 2 hours will have no significant impact on

probability or consequences of any accident previously evaluated.

When an RTB is inoperable, Functional Testing and other breaker

operations becomes more difficult. The current technical

specification allows an inoperable breaker to be closed for 1 hour

to perform testing of other RTBs. This provision is infrequently

required, but when it is required, the allowed outage time is very

short and rushing to complete a test may lead to an inadvertent

reactor trip. Increasing this allowed outage time is an improvement

item identified in NUREG 1366 and consistent with philosophy

provided in Generic Letter 89-07.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

This proposed change does not involve any changes in equipment

and will not alter the manner in which the plant will be operated.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

The proposed change will not adversely affect the performance of

the safety function of the RTBs. In fact, it is expected that the

performance of the RTBs will improve as a result of this change

based on less wear and tear on the equipment. The proposed change

will have no adverse impact on the protective boundaries, safety

limits or margin of safety.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

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

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

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

NRC Project Director: John N. Hannon.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic City Electric Company, Docket No.

50-277, Peach Bottom Atomic Power Station, Unit No. 2, York County,

Pennsylvania

Date of application for amendment: July 10, 1998.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to incorporate revised Safety

Limit Minimum Critical Power Ratios (SLMCPRs) for the use of cycle-

specific analysis performed for Peach Bottom Atomic Power Station

(PBAPS), Unit 2, Cycle 13.

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.

[[Page 48262]]

The derivation of the cycle-specific SLMCPRs for incorporation

into the TS, and its use to determine cycle-specific thermal limits,

have been performed using the methodology discussed in ``General

Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13,

and U.S. Supplement, NEDE-24011-P-A-13-US, August, 1996, and the

``Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-

A (GESTAR II) on Cycle Specific Safety Limit MCPR.'' Amendment 25

was submitted by [General Electric Nuclear Energy] GENE to the U.S.

Nuclear Regulatory Commission (USNRC) on December 13, 1996. This

change in SLMCPRs cannot increase the probability or severity of an

accident.

The basis of the SLMCPR calculation is to ensure that greater

than 99.9% of all fuel rods in the core avoid transition boiling if

the limit is not violated. The new SLMCPRs preserve the existing

margin to transition boiling and fuel damage in the event of a

postulated accident. The fuel licensing acceptance criteria for the

SLMCPR calculation apply to PBAPS, Unit 2, Cycle 13 in the same

manner as they have applied previously. The probability of fuel

damage is not increased. Therefore, the proposed TS changes do not

involve an increase in the probability or consequences of an

accident previously evaluated.

In addition to the change to the SLMCPR, the footnote to TS

2.1.1.2 is being revised, and a footnote is being added to TS

5.6.5.b.1. The revision to the footnote associated with TS 2.1.1.2

will ensure that the SLMCPR value is reconfirmed for the cycle

subsequent to PBAPS, Unit 2, Cycle 13, and the footnote to TS

5.6.5.b.1 is being added due to the use of the proposed Amendment 25

and the use of a proposed R-factor calculation methodology (``R-

Factor Calculation Method for GE11, GE12, and GE13 Fuel,'' NEDC-

32505P, Revision 1, June 1997), which has not yet been approved for

generic use by the USNRC. The revision to the footnote associated

with TS 2.1.1.2 and the addition of the footnote to TS 5.6.5.b.1 are

administrative changes that do not involve an increase in the

probability or 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 SLMCPR is a TS numerical value, designed to ensure that

transition boiling does not occur in 99.9% of all fuel rods in the

core during the limiting postulated accident. It cannot create the

possibility of any new type of accident. The new SLMCPRs are

calculated using methodology discussed in ``Generic Electric

Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, and U.S.

Supplement, NEDE-24011-P-A-13-US, August, 1996, and the ``Proposed

Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR

II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 was

submitted by GENE to the USNRC on December 13, 1996. Therefore, the

revision to the SLMCPR will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Additionally, this proposed change will revise the footnote to

TS 2.1.1.2, and add a footnote to TS 5.6.5.b.1. The revision to the

footnote associated with TS 2.1.1.2, and the addition of the

footnote to TS 5.6.5.b.1, are administrative changes that do not

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

any previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

There is no significant reduction in the margin of safety

previously approved by the USNRC as a result of the proposed change

to the SLMCPR, and the proposed change that will revise the footnote

to TS 2.1.1.2, and add a footnote to TS 5.6.5.b.1. The new SLMCPRs

are calculated using methodology discussed in ``General Electric

Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, and U.S.

Supplement, NEDE-24011-P-A-13-US, August, 1996, and the ``Proposed

Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR

II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 was

submitted by GENE to the USNRC on December 13, 1996. The fuel

licensing acceptance criteria for the calculation of the SLMCPR

apply to PBAPS, Unit 2 Cycle 13 in the same manner as they have

applied previously. The SLMCPRs ensure that greater than 99.9% of

all fuel rods in the core will avoid transition boiling if the limit

is not violated, thereby preserving the fuel cladding integrity.

Therefore, the proposed TS changes will not significantly reduce the

margin of safety previously approved by the USNRC.

Additionally, the proposed change that will revise the footnote

to TS 2.1.1.2, and add a footnote to TS 5.6.5.b.1 is an

administrative change that will not significantly reduce the margin

of safety previously approved by the USNRC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (Regional Depository) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105.

Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101.

NRC Project Director: Robert A. Capra.

Pennsylvania Power and Light Company, Docket No. 50-388 Susquehanna

Steam Electric Station, Unit 2, Luzerne County, Pennsylvania.

Date of amendment request: August 4, 1998

Description of amendment request: The amendment would modify the

Susquehanna Steam Electric Station, Unit 2, Technical Specifications to

replace figures 2.1.1.2-1 and 2.1.1.2-2, and associated footnotes, with

single value minimum critical power ratio (MCPR) Safety Limits of

Section 2.1.1.2; remove references from Section 5.6.5 which do not

directly support the generation of Core Operating Limits; remove

references from Section 5.6.5 which were previously included to address

the application of the ANFB-10 correlation to ATRIUM-10 fuel; include

Siemens Power Corporation ANFB-10 topical report in Section 5.6.5; and

to change the Bases to reflect inclusion of the ANFB-10 critical power

correlation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The applicable sections of the FSAR [Final Safety Analysis

Report] are Chapters 4.4 and 15. FSAR Chapter 4.4 describes the MCPR

Safety Limit, and Chapter 15 describes the transient and accident

analyses. The reference to be added to Section 5.6.5 of the Unit 2

Technical Specifications describes an NRC approved critical power

correlation for ATRIUMTM-10 fuel appropriate for use in

conservative methodologies for generating MCPR Safety Limits and

MCPR Operating Limits to assure safe operation of Unit 2 with

ATRIUMTM-10 fuel. A discussion of the impact of the

proposed Technical Specification change is provided below.

The proposed change in critical power correlation does not

physically affect the plant or its systems. Thus, it does not

increase the probability of an accident previously evaluated.

A Unit 2 Cycle 10 MCPR Safety Limit analysis was performed for

PP&L by SPC. This analysis used NRC approved methods described in

ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2. These methods

will be used each cycle to calculate the Unit 2 Safety Limits. For

Unit 2 Cycle 10, the critical power performance of the 9x9-2 and

ATRIUMTM-10 fuel was determined using the NRC approved

ANFB and ANFB-10 correlations, respectively. The SAFETY LIMIT MCPR

calculations statistically combine uncertainties on feedwater flow,

feedwater temperature, core flow, core pressure, core power

distribution, and uncertainties in the Critical Power Correlations.

The SPC analysis used cycle specific power distributions and

calculated MCPR values such that at least 99.91% of the fuel rods

are expected to avoid boiling transition during normal operation or

anticipated operational occurrences. The resulting two-loop and

single-loop MCPR Safety Limits are included in the proposed

Technical Specification change. Thus, the cladding integrity and its

ability to contain fission products are not adversely affected.

[[Page 48263]]

Analyses of the Single Loop Pump Seizure accident with the NRC

approved ANFB-10 correlation for the ATRIUMTM-10 fuel

(Reference 1) [Reference 1 refers to the reference listed in the

application dated August 4, 1998] will be performed to demonstrate

that the NRC acceptance criterion (i.e., small fraction of 10 CFR

100 dose limits) is met. Analyses will also be performed to validate

the conclusion that single-loop transients are less severe than the

those events analyzed for two-loop operation.

Changes to Section 2.1.1.2 reflect the change from a flow

dependent MCPR Safety Limit to a single value MCPR Safety Limit for

two-loop operation and single-loop operation.

Changes to Reference 5.6.5 delete the methodology used for

critical power analyses for ATRIUMTM-10 fuel and add the

NRC approved ANFB-10 methodology to the list of approved

methodologies. Other changes in Reference 5.6.5 are administrative

in nature because they delete references that are not directly

related to the generation of Core Operating Limits. No new analysis

approaches are used due to the removal of these references.

Changes to BASES Sections 2.1.1 and 3.2.2 reflect the inclusion

of the ANFB-10 critical power correlation. The range of the

applicability of the ANFB-10 is valid for pressures > 571 psia and

bundle mass fluxes > 0.115 x 10 \6\ lb/hr-ft \2\. These values

assure that a valid CPR calculation will result at or above 25% of

rated core thermal power, that is, reactor steam dome pressure

785 psig and core flow 10 Mlbm/hr.

The consequences of transients and accidents will remain within

the criteria approved by the NRC. The methodology used to perform

the analyses have been previously approved by the NRC. Thus,

analysis results using the new methodology will continue to provide

assurance that the reactor will perform its design safety function

during normal operation and design basis events. Therefore, the

proposed action does not involve an increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes to the Unit 2 Technical Specifications

(MCPR Safety Limits, removal of methodology references not directly

supporting the generation of Core Operating Limits, removal of the

two references describing previously approved methodology for

applying ANFB to ATRIUMTM-10 fuel, and inclusion of the

ANFB-10 correlation reference) do not require any physical plant

modifications, physically affect any plant components, or entail

changes in plant operation. Removal of the Unit 2 Cycle 9 footnote

allows Unit 2 Cycle 10 and future cycle operation with thermal

limits generated using NRC approved methodology. Thus, the proposed

change does not create the possibility of a previously unevaluated

operator error or a new single failure. The consequences of

transients and accidents will remain within the criteria approved by

the NRC. 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 applicable Technical Specification Sections include 2.1.1.2

and 5.6.5.

The changes to the Unit 2 Technical Specifications discussed in

item 1 above do not require any physical plant modifications,

physically affect any plant components, or entail changes in plant

operation. Therefore, the proposed change will not jeopardize or

degrade the function or operation of any plant system or component

governed by Technical Specifications. The consequences of transients

and accidents will remain within the criteria approved by the NRC.

The proposed MCPR Safety Limits and use of the NRC approved ANFB-10

critical power correlation described in the reference added to

Section 5.6.5 do not involve a significant reduction in the margin

of safety as currently defined in the BASES of the applicable

Technical Specification sections.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: Robert A. Capra.

Pennsylvania Power and Light Company (PP&L), Docket No. 50-388,

Susquehanna Steam Electric Station, Unit 2, Luzerne County,

Pennsylvania

Date of amendment request: August 5, 1998.

Description of amendment request: The amendment would modify the

Susquehanna Steam Electric Station, Unit 2, Technical Specifications

Table 3.3.5.1-1 ``Emergency Core Cooling System Instrumentation.'' The

change updates the allowable values for both the Core Spray (CS) and

Low Pressure Coolant Injection System (LPCI) ``Reactor Steam Dome

Pressure--Low'' functions for initiation and injection permissive.

Specifically, the allowable values are changed from a specified minimum

pressure to a specified allowable pressure band. This more restrictive

allowable value range will prevent CS and LPCI system

overpressurization while still permitting injection to prevent fuel

clad temperature limits from being exceeded.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This proposal does not involve an increase in the probability or

consequences of an accident previously evaluated. The proposed

amendment changes the ``Reactor Steam Dome Pressure--Low'' Allowable

Values so to provide further assurance that the Core Spray and

[Residual Heat Removal] RHR systems will perform their [loss-of-

coolant accident] LOCA design basis function.

The functional design basis of the Core Spray and LPCI is to

inject water into the reactor vessel to cool the core during a LOCA

by opening the Core Spray and LPCI injection valves when reactor

pressure drops below the reactor vessel low pressure permissive. The

upper analytical limit for the permissive is the Core Spray and LPCI

systems' maximum design pressure, and the lower analytical limit is

the lowest pressure which allows injection to prevent exceeding the

fuel cladding temperature limit. The new allowable values were

selected to lie within the upper and lower limits to ensure there

will be no change in the required logic or functions of the Core

Spray and LPCI systems. These new values do not affect the LOCA nor

its ``limiting fault'' frequency of occurrence and do not introduce

any new accidents or malfunctions of equipment important to safety.

Since they do not affect the LOCA, they do not change the

probability of occurrence of the LOCA. The new allowable values do

not change the logic or function of the reactor vessel low pressure

permissive. These new values simply provide the basis for which the

associated pressure instruments are to be set to ensure proper

operation of Core Spray and LPCI within the design pressures as

described above. Therefore, the change in allowable values does not

increase the probability of occurrence or the consequences of an

accident or malfunction of equipment important to safety.

Based upon the analysis presented above, PP&L concludes that the

proposed action does not involve an increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This proposal does not create the probability of a new or

different type of accident from any accident previously evaluated.

The new allowable values do not change any plant systems,

structures, or components, nor do they change any existing or create

any new Core Spray and LPCI logic or functions. The new allowable

values were selected to ensure the required operation of the Core

Spray and LPCI systems within the maximum design pressures.

[[Page 48264]]

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The change does not involve a reduction in the margin of safety.

Technical Specification Bases Section B3.3.5.1 [9 sic] (ECCS

Instrumentation) identifies that the low reactor steam dome pressure

signals are used as permissives for operation of the low pressure

ECCS subsystems. The new allowable values were selected so as to not

impact the logic, redundancy, operability or surveillance

requirements for these subsystems. The new allowable values maintain

the margin requirements of the Core Spray and LPCI system pressures

such that they do not exceed their system maximum design pressures

and that system pressures are high enough to ensure that the ECCS

injection prevents the fuel peak cladding temperature from exceeding

the limits of 10 CFR50.46.

Therefore, the margin of safety is enhanced by the proposed

changes.

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: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: Robert A. Capra.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: August 3, 1998.

Description of amendment request: The proposed changes provide for

applicability of the safety limit minimum critical power ratio (SLMCPR)

to fuel 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:

Operation of the FitzPatrick plant in accordance with the proposed

amendment would not involve a significant hazards consideration as

defined in 10 CFR 50.92, since it would not:

1. involve a significant increase in the probability or

consequences of an accident previously evaluated.

A change to a note stating that the SLMCPR remains applicable

through Cycle 14 does not affect the initiation of any accident.

Operation in accordance with the current SLMCPR ensures the

consequences of previously analyzed accidents are not changed.

Therefore, this proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

from any accident previously evaluated.

The SLMCPR establishes a performance limit for the fuel. This

limit remains unchanged. Changing a note to reflect this is an

administrative change and will not initiate any accident. Therefore,

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

different kind of accident from any accident previously evaluated.

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

GE [General Electric] has performed an evaluation of the SLMCPR

for Cycle 14 and found that the cycle specific value, based on

current reload plans, is bounded by the generic value calculated for

GE 12 fuel. The existing SLMCPR remains unchanged for Cycle 14 and

the margin of safety for the prevention of onset of transition

boiling is unchanged. Therefore, this proposed change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

York, New York 10019.

NRC Project Director: S. Singh Bajwa, Director.

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

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

Jersey

Date of amendment request: July 30, 1998.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3/4.7.6, ``Control Room Emergency

Air Conditioning System.'' Specifically, the acceptance criteria for

the control room envelope would be revised to maintain a \1/8\-inch

positive pressure with respect to all areas directly accessible from

the control room and a positive pressure with respect to all other

areas adjacent to the control room.

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.

CREACS [Control Room Emergency Air Conditioning System] ensures

adequate protection after an accident and is not an accident

initiator. The change to the acceptance criteria for CREACS does not

affect the probability of an accident.

Revising the acceptance criteria for the CREACS from a `1/8-inch

W.G. [water gauge] positive pressure in the control room with

respect to the adjacent area' to `a 1/8-inch W.G. positive pressure

in the control room with respect to all areas directly accessible

(Work Control Center and Control Room Equipment Rooms) from the

control room and a positive pressure to all other areas adjacent to

the control room' does not alter the assumptions in the radiological

dose assessment provided to the NRC and approved under Amendments

190 (Unit 1) and 173 (Unit 2). Therefore the conclusions of the

radiological dose assessment reviewed and approved by the NRC under

the above Amendments remain unchanged. The radiological dose

assessment provided under Amendments 190 and 173 demonstrates that

operation of the CREAS in the pressurized mode at the initiation of

an accident will ensure that the requirements of General Design

Criterion (GDC) 19 will be met.

Therefore, the proposed TS 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.

Since the CREACS is an accident mitigation system that does not

communicate with the Reactor Coolant Pressure boundary or interface

with Emergency Core Cooling Systems (ECCS), the proposed change to

the acceptance criteria for CREACS pressurization cannot result in

new accident scenarios. The function of the CREACS system is to

maintain the habitability of the CRE [control room envelope]

following an accident.

Therefore, the proposed TS change does not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The CREACS ensures that (1) the ambient air temperature does not

exceed the allowable temperature for continuous duty rating for

equipment and instrumentation cooled by the CREACS and (2) the

Control Room will remain habitable for operations personnel during

and following all credible radiological accident conditions.

Revising the

[[Page 48265]]

acceptance criteria to maintaining the control room at a \1/8\-inch

W.G. positive pressure in the control room with respect to all areas

directly accessible (Work Control Center and Control Room Equipment

Rooms) from the control room and a positive pressure to all other

areas adjacent to the control room does not alter the assumptions

used in the radiological dose assessment nor revise the conclusions

of the dose assessment which was reviewed under Amendments 190 and

173. Since the assumptions and conclusions of the dose assessment

remain unchanged, the CREACS continues to ensure that the

requirements of GDC 19 continue to be met, and there is no reduction

in the safety provided to the control room operators.

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

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079.

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

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

NRC Project Director: Robert A. Capra.

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: August 12, 1998.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3/4.6.1.3, ``Containment Air

Locks,'' to change the action statements for an inoperable airlock. The

proposed amendments would also correct an editorial error in TS Bases

3/4.6.1.2, ``Containment Leakage.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The reactor containment serves to mitigate the consequences of a

Design Basis Accident (DBA). That is, the containment is designed to

provide a barrier to ensure that in the event of a DBA, a release of

radioactive material will not result in the radiation dose to the

general public exceeding the limits of 10 CFR 100. Each unit's

containment has been provided with two air locks. These air locks

permit personnel to access components and systems within the

containment boundary without compromising the containment's ability

to carry out its design function. In this capacity, the air locks

serve as part of the containment boundary and as such are not

considered as a contributor to the probability of an accident.

To carry out their design function, the air locks are designed

and tested to certify their ability to withstand a pressure in

excess of the maximum expected following a DBA. Each door is

individually tested to verify that leakage will remain below design

values with the containment at design pressure. An interlock is

provided to ensure that containment integrity is maintained during

personnel passage by allowing only one air lock door to be open at a

time. This interlock is also periodically tested to verify its

functionality.

The proposed changes will allow continued operation with one air

lock door inoperable or with the air lock door interlock mechanism

disabled but will specify the actions necessary under those

conditions to assure that containment integrity is not compromised.

This will ensure that the consequences of an accident previously

evaluated are not significantly increased. Additionally, the

proposed changes specify that in the event that an air lock is

inoperable for a reason other than an inoperable air lock door, or

air lock interlock mechanism, the unit must be placed in a condition

in which the analyzed accident could not occur.

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

significant increase in the probability or consequences of an

accident previously evaluated.

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

accident.

The proposed changes to the Containment Air Lock Technical

Specifications do not affect the ability of the containment to carry

out its design function. The changes also do not introduce any new

equipment; nor do they result in the operation of the plant in a

manner contrary to the safety analysis. Therefore, the proposed

changes will not increase the probability of a new or different kind

of accident from any accident previously identified.

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

safety.

The proposed changes do not affect any design or functional

requirements of the Containment or the Containment Air Locks.

Additionally, the proposed changes do not affect any of the

conditions or assumptions of the applicable safety analyses.

Containment Air Lock leakage rates are determined based upon

containment leakage at design pressure. The proposed changes will

not affect containment design pressure nor will they affect the peak

containment pressures expected for analyzed accidents.

Based upon the above, the proposed change will 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. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079.

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

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

NRC Project Director: Robert A. Capra.

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

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

Diego County, California

Date of amendment requests: May 11, 1998 (Supersedes the May 30,

1996, amendment request). This Notice supersedes the staff's proposed

no significant hazards consideration determination for the requested

changes that was published on September 11, 1996 (61 FR 47981).

Description of amendment requests: The proposed amendments would

revise the Technical Specifications (TS) to allow use of performance-

based criteria to establish containment leak rate test intervals and

add a new ``Containment Leakage Rate Testing Program'' to the

administrative section of TS to codify the program used to determine

the testing program. The proposed program implements 10 CFR Part 50,

Appendix J, Option B, by referring to Regulatory Guide 1.163,

``Performance-Based Containment Leak Test Program,'' dated September

1995.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Since the interval between containment leakage rate tests is not

related in any way to conditions which cause accidents, and plant

structures, systems, and components will not be operated in a

different manner as a result of the proposed Technical Specification

(TS) change, the proposed changes will not increase the probability

of an accident previously evaluated.

Containment leakage may result from accidents which are

evaluated in the Updated Final Safety Analysis Report. The proposed

TS changes may result in an acceptably small increase in post-

accident containment leakage. Using a statistical approach, NUREG-

1493 determined that the increase in hypothetical dose to the public

resulting from extending the testing interval is extremely small.

NUREG-1493 concluded that such small hypothetical dose increases

[[Page 48266]]

to the public are justifiable due to the real reduction in

occupational exposure resulting from interval extension. Therefore,

the proposed change does not significantly increase the consequences

of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change only incorporates the performance based

approach for containment leak rate testing authorized in the new

Option B to Appendix J of 10 CFR Part 50. The interval extensions

allowed, through this approach, do not have the potential for

creating the possibility of new or different kinds of accidents from

those previously evaluated because plant structures, systems, and

components will not be operated in a different manner as a result of

the TS change and, therefore, will not introduce any new or

different failure modes or initiators. Therefore, the proposed

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

accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed Technical Specification does not alter the

allowable containment leakage rate. The proposed change replaces the

current, prescriptive testing requirements with a new performance

based approach for establishing the testing intervals. Therefore,

the proposed change does not involve a significant reduction in a

margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

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

California Edison Company, P. O. Box 800, Rosemead, California 91770.

NRC Project Director: William H. Bateman.

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

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

Diego County, California

Date of amendment requests: June 19, 1998.

Description of amendment requests: The proposed amendments would

revise Technical Specification (TS) 3.4.1, ``RCS DNB (Pressure,

Temperature and Flow) Limits.'' Specifically, the proposed changes

would include (1) a reduction in the minimum primary reactor coolant

system (RCS) cold leg temperature (Tcold) from 554 F to 535

F between the 70 percent and 100 percent rated thermal power levels,

(2) a conversion of the specified RCS minimum flow rate from a ``Mass''

(i.e., lb/hr) to a ``Volumetric'' (gpm) flow basis, and (3) elimination

of the maximum RCS flow rate limit.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change to Technical Specification (TS) 3.4.1 does

not adversely impact structure, system, or component design or

operation in a manner which would result in a change in the

frequency of occurrence of accident initiation. Nor are the affected

parameters themselves accident initiators. As such, the proposed TS

change will not significantly increase the probability of accidents

previously evaluated. Likewise, the proposed TS change does not

significantly increase the consequences of an accident previously

evaluated. The safety analysis assessments confirm that the existing

Analyses of Record (AORs) for San Onofre Units 2 and 3 remain valid

or have been re-analyzed to demonstrate continued compliance with

applicable Acceptance Criteria.

The change in Reactor Coolant System (RCS) ``Mass'' flow to

``Volumetric'' flow is a change in measuring units to be consistent

with the measure used in the performance of the safety analysis.

Therefore, there is no impact on any evaluated accidents.

The elimination of the upper RCS flow limit has no effect on

Departure from Nucleate Boiling which is a concern at lower flows,

and the maximum flow that is physically possible is less than the

current upper limit.

Therefore, this amendment request 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.

Tcold is an input parameter used in event analysis,

it is not an event initiator. No new or different accidents have

been identified which could result from operating at the proposed

Tcold. The safety analysis assessments performed confirm

that the existing safety system settings for San Onofre Units 2 and

3 remain valid, thereby assuring continued conformance to the

Acceptance Criteria for all events.

A change in RCS flow measuring units can not initiate an

accident, nor can the elimination of an upper RCS flow limit which

can not be attained.

Therefore, this amendment 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.

Updated Final Safety Analysis Report (UFSAR) safety analyses

have been assessed and remain valid or have been re-analyzed to

demonstrate continued compliance with applicable Acceptance Criteria

for operation at the reduced Tcold. All other safety

limits and safety system settings remain unchanged.

A change in measuring units for RCS flow does not reduce the

margin of safety.

Elimination of an RCS flow limit that can not physically be

reached does not reduce the margin of safety. The shiftly

surveillance requirement for maximum flow has no practical basis or

safety benefit. Additionally, the margin to departure from nuclear

boiling increases as the flow rate increases.

Therefore, this amendment does not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

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

California Edison Company, P.O. Box 800, Rosemead, California 91770.

NRC Project Director: William H. Bateman.

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

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

Date of amendment request: July 6, 1998.

Description of amendment request: The proposed amendment would

relocate the description of the reactor coolant system design features

from Technical Specification 5.4 to the Updated Final Safety Analysis

Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change relocates the description of the Reactor

Coolant System design features to the Updated Final Safety Analysis

Report (UFSAR), a licensee-controlled document. The description of

the Reactor Coolant System design features, currently a part of the

UFSAR, is maintained in accordance with 10 CFR 50.59 and 50.71.

Existing South Texas Project procedures ensure that changes to the

facility as described in the UFSAR, such as the replacement of the

steam generators, are reviewed to determine if an unreviewed

[[Page 48267]]

safety question exists. The proposed amendment does not result in

any hardware or operating procedure changes. The initiators of any

accident previously evaluated are not affected by the relocation of

the Reactor Coolant System design features. 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.

The change does not alter the plant configuration or make

changes in the methods governing plant operation. The proposed

change does not impose different requirements, and adequate control

of information will be maintained in accordance with existing

procedures. The change does not alter assumptions made in the safety

analysis and licensing basis. Therefore, the proposed change does

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

3. The proposed change does not involve a significant reduction

in a margin of safety.

The relocation of a description of Reactor Coolant System design

features has no impact on any safety analysis assumptions. There are

no changes to the plant configuration or operating procedures.

Future changes to the relocated information are governed by existing

procedures in accordance with 10 CFR 50.59 and 50.71. Consequently,

there is no significant reduction in a margin of safety.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J.M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

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

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: July 6, 1998.

Description of amendment request: Relocates the Technical

Specification 3/4.3.3.3 requirements for the Seismic Instrumentation to

the Technical Requirements Manual.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relocates requirements and surveillances for

the Seismic Monitoring System that do not meet the criteria for

inclusion in Technical Specifications as identified in 10 CFR

50.36(c)(2)(ii). The affected systems and components are not assumed

to be initiators of analyzed events and are not assumed to mitigate

accident or transient events. The requirements and surveillances for

these affected systems and components will be relocated from the

Technical Specifications to the Technical Requirements Manual, which

is incorporated in the STP UFSAR and will be maintained pursuant to

10 CFR 50.59. In addition, the Seismic Monitoring System components

are addressed in existing surveillance procedures which are also

controlled by 10 CFR 50.59 and subject to the change control

provisions imposed by plant administrative procedures, which endorse

applicable regulations and standards. The associated changes to the

Index are administrative. Therefore, the change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change relocates requirements and surveillances for

the Seismic Monitoring System that do not meet the criteria for

inclusion in Technical Specifications as identified in 10 CFR

50.36(c)(2)(ii). The change does not involve a physical alteration

of the plant (no new or different type of equipment will be

installed) or make changes in the methods governing normal plant

operation. The change will not impose different requirements, and

adequate control of information will be maintained. This change will

not alter assumptions made in the safety analysis and licensing

basis. The associated changes to the Index are administrative.

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

different kind of accident from any accident previously evaluated.

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

of safety?

The proposed change relocates requirements and surveillances for

the Seismic Monitoring System, which does not meet the 10 CFR 50.36

criteria for inclusion in Technical Specifications. The change will

not reduce a margin of safety because the change has no impact on

any safety analysis assumptions. In addition, the relocated

requirements and surveillances for the affected structures, systems,

components, or variables remain the same as the existing Technical

Specifications. Since any future changes to these requirements or

the surveillance procedures will be evaluated per the requirements

of 10 CFR 50.59, there will be no reduction in a margin of safety.

The associated changes to the Index are administrative and have no

potential effect on the margin of safety.

Therefore, the change does not involve a significant reduction

in the margin of safety.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

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

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: July 6, 1998.

Description of amendment request: Relocates the Technical

Specification 3/4.7.13 requirements for the Area Temperature Monitoring

System to the Technical Requirements Manual.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relocates requirements and surveillances for

Technical Specification 3/4.7.13, which does not meet the criteria

for inclusion in Technical Specifications as identified in 10 CFR

50.36(c)(2)(ii). The affected systems and components are not assumed

to be initiators of analyzed events and are not assumed to mitigate

accident or transient events. The requirements and surveillances for

these affected systems and components will be relocated from the

Technical Specifications to the Technical Requirements Manual, which

is incorporated in the STP UFSAR and will be maintained pursuant to

10 CFR 50.59. In addition, the Area Temperature Monitoring System

components are addressed in existing surveillance procedures which

are also controlled by 10 CFR 50.59 and subject to the change

control provisions imposed by plant administrative procedures, which

endorse applicable regulations and standards. The associated changes

to the Index are administrative. Therefore, the change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

[[Page 48268]]

The proposed change relocates requirements and surveillances for

the Area Temperature Monitoring System, which does not meet the

criteria for inclusion in Technical Specifications as identified in

10 CFR 50.36(c)(2)(ii). The change does not involve a physical

alteration of the plant (no new or different type of equipment will

be installed) or make changes in the methods governing normal plant

operation. The change will not impose different requirements, and

adequate control of information will be maintained. This change will

not alter assumptions made in the safety analysis and licensing

basis. The associated changes to the Index are administrative.

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

different kind of accident from any accident previously evaluated.

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

of safety?

The proposed change relocates requirements and surveillances for

the Area Temperature Monitoring System, which does not meet the 10

CFR 50.36 criteria for inclusion in Technical Specifications. The

change will not reduce a margin of safety since it has no impact on

any safety analysis assumptions. In addition, the relocated

requirements and surveillances for the affected structure, system,

component, or variable remain the same as the existing Technical

Specifications. Since any future changes to these requirements or

the surveillance procedures will be evaluated per the requirements

of 10 CFR 50.59, there will be no reduction in a margin of safety.

The associated changes to the Index are administrative and have no

potential effect on the margin of safety.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

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

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: July 22, 1998.

Description of amendment request: The proposed amendment would

revise the Technical Specifications to reflect the steam generator

water level low-low trip setpoint differences between the existing

Model E and the replacement Model Delta-94 steam generators for the

Reactor Trip System and the Engineered Safety Features Actuation System

instrumentation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This proposed change includes changing the low-low steam

generator water level trip setpoint. The setpoint is being changed

to enhance the operational flexibility associated with the RSGs

[replacement steam generators].

The minimum setpoint change proposed in this request establishes

controls to ensure that an adequate heat sink is maintained by

providing an adequate secondary liquid mass to remove primary system

sensible heat and core decay heat shortly after reactor trip and

initiating auxiliary feedwater flow for long-term cooling. The

accidents analyzed for this requirement are the Loss of Non-

Emergency AC Power to the Plant Auxiliaries, Loss of Normal

Feedwater and Feedwater Line Break transients. These accidents were

analyzed utilizing the Westinghouse RETRAN model. All acceptance

criteria were shown to be met for both these events. Therefore, the

proposed steam generator water level low-low trip setpoint change is

demonstrated not to result in an increase in the consequences for

these accidents.

The steam generator water level low-low trip setpoint is not

considered a precursor to any of the analyzed accidents, and

therefore, these proposed changes do not result in an increase in

the probability or consequences of any accident previously analyzed.

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

or different kind of accident from any accident previously

evaluated.

The proposed setpoint change does not create any new operating

conditions or modes. The proposed change only revises the actuation

setpoints for the Reactor Trip System and Engineered Safety Features

Actuation System. The actions of these systems continue to be

performed in accordance with existing requirements, which are

sufficient to ensure plant safety is maintained.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The events potentially affected by the setpoint change in the

steam generator water level low-low reactor trip (Table 2.2-1,

Function 13) and ESFAS Auxiliary Feedwater System actuation (Table

3.3-4, Function 6.d) are the Loss of Normal Feedwater and Feedwater

System Pipe Break. These events were analyzed and it was

demonstrated that all acceptance criteria were met for both of these

events.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

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

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: July 28, 1998.

Description of amendment request: The proposed amendment addresses

the operator action to reduce the steam generator power-operated relief

valve setpoint consistent with the revised small-break loss-of-coolant

accident (SBLOCA) analysis for the replacement Delta-94 steam

generators. The operator action and the associated revised SBLOCA

analysis are reflected in a proposed revision to the South Texas

Project Updated Final Safety Analysis Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed operator action associated with the re-analysis of

the Delta-94 SGs [steam generators] will not result in a significant

increase in the probability of an accident previously evaluated. The

initiators of any design basis accident are not affected by this

operator action. The operator action would facilitate the automatic

mitigation capability of the SG PORVs [power-operated relief

valves], and would not initiate the mitigating safety function. The

operator action will be incorporated into the EOPs [Emergency

Operating Procedures] and would not be performed until after the

initiation of an accident. The automatic actuation of the SG PORVs

is not a new design feature. The effects of inadvertent opening of a

single steam dump, relief or safety valve are currently analyzed as

described in Section 15.1.4 of the UFSAR [Updated Final Safety

Analysis Report]. Consequently, there is no significant impact on

any previously evaluated accident probabilities.

[[Page 48269]]

The proposed operator action associated with the re-analysis of

the Delta-94 SGs does not result in a significant increase in the

consequences of any accidents previously evaluated. The operator

action will not adversely affect the integrated ability of the plant

systems to perform their intended safety functions to mitigate the

consequences of a small break LOCA [loss-of-coolant accident], or

any other accident previously evaluated. In fact, the re-analysis

has demonstrated that the use of the operator action reduces the

consequences of a small break LOCA in that the Peak Cladding

Temperature for the most limiting small break LOCA transient is

reduced and continues to be substantially below the acceptance limit

of 10 CFR 50.46.

The operator action does not affect the integrity of any fission

product barrier such that their function in the control of

radiological consequences is not affected. The radiological

consequences for the small break LOCA presented in the UFSAR remain

unchanged as a result of the proposed operator action.

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

or different kind of accident from any accident previously

evaluated.

The proposed license amendment does not create the possibility

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

evaluated. The proposed amendment is not the result of any physical

changes to the existing facility. The operator action does not

represent a different initiator for any design basis accident and

does not create new design basis scenarios. Small break LOCA

mitigation, utilizing a combination of automatic and manual actions,

is already part of the STP [South Texas Project, Units 1 and 2]

licensing basis. Written procedures address those operator actions

required for small break LOCA mitigation. The current STP EOPs have

an operator action for a steam generator tube rupture (SGTR) similar

to the operator action for the small break LOCA addressed by this

proposed license amendment. The operator action for the SGTR is to

raise the safety-grade SG PORV setpoints. The operator action

credited in the small break LOCA analysis for the Delta-94 SGs is to

lower the safety-grade SG PORV setpoints. The purpose of the action

is to provide a more rapid cooldown of the primary side by

depressurizing the secondary side during a small break LOCA using

the steam dumps first, then the SG PORVs, if steam dumps are

unavailable. The inadvertent operation of a single steam dump,

relief or safety valve is currently addressed in UFSAR Section

15.1.4.

The proposed amendment does not alter any original design

specification, such as seismic requirements, electrical separation

requirements and environmental qualification, and is not the result

of any physical changes to the facility. In addition, the proposed

amendment does not result in exposure of additional equipment used

in accident mitigation to an adverse environment beyond that

currently identified in the UFSAR.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed operator action does not involve a significant

reduction in the margin of safety. The plant systems required for

the mitigation of any design basis accidents will continue to be

able to perform their safety function. In fact, the re-analysis has

demonstrated that the use of the operator action reduces the

consequences of a small break LOCA in that the Peak Cladding

Temperature for the most limiting small break LOCA transient is

reduced and continues to be substantially below the acceptance

criteria of 10 CFR 50.46.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

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

NRC Project Director: John N. Hannon.

Tennessee Valley Authority, Docket No. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant Units 1, 2, 3, Limestone County, Alabama

Date of amendment request: June 12, and August 14, 1998.

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) for the Browns Ferry Nuclear

Plant (BFN) Units 1, 2 and 3. The proposed changes would revise

surveillance frequency of ``once-per-cycle'' surveillance requirements

(SR) from 18 to 24 months to accommodate a 24-month fuel cycle. The

licensee also proposed changes to the associated TS Bases (TS-390).

Basis for proposed no significant hazards consideration

determination: Tennesee Valley Authority addressed the affected SRs

into two groups: (1) non-instrument calibration related, and (b) those

involving instrument calibrations. 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:

Group 1: Non-instrument Calibration Related SRs

(1) The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment changes the surveillance frequency from

18 months to 24 months for SRs in the Units 2 and 3 TS that are

normally a function of the refueling interval. In addition, the

proposed amendment changes the surveillance frequency from 18 months

to 24 months for those SRs in the Unit 1 TS that control the test

interval for components and systems that are common to Units 1, 2,

and 3. Under certain circumstances SR 3.0.2 would allow a maximum

surveillance interval of 30 months for these SRs. The evaluations in

Section III [Licensee's June 12, 1998 application, Section III,

Safety Analysis] have shown that the reliability of protective

instrumentation and equipment will be preserved for the maximum

allowable surveillance interval. The proposed changes do not involve

any change to the design or functional requirements of plant

systems, and the surveillance test methods will be unchanged. The

proposed changes will not give rise to any increase in operating

power level, fuel operating limits, or effluents. In addition, the

proposed changes will not significantly increase any radiation

levels. Based on the foregoing considerations and the evaluations

completed in accordance with the guidance of Generic Letter 91-04,

it is concluded that the proposed amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed amendment requires no change to the plant design or

the mode of operation, for any item of equipment. No new equipment

is either added or substituted for any existing equipment. Based on

the Section III [Licensee's June 12, 1998 application, Section III,

Safety Analysis] evaluations, the extension of surveillance

intervals is shown to have no significant impact on equipment

performance. The proposed changes do not create the possibility of

any new failure mechanisms. Therefore, the proposed amendment does

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

from any accident previously evaluated.

(3) The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed amendment seeks to change surveillance intervals

from 18 to 24 months. Although the proposed TS changes will result

in an increase in the interval between surveillance tests, the

impact on system availability is small based on other, more frequent

testing or redundant systems or equipment. There is no evidence of

any failures that would impact the availability of the systems. This

change does not alter the existing setpoints, TS allowable values or

analytical limits. The assumptions in the current safety analyses

are not impacted and the proposed amendment does not reduce a margin

of safety.

Therefore, it is concluded that the proposed amendment does not

involve a significant reduction in a margin of safety.

Group 2: SRs that Involve Instrument Calibrations

(1) The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

[[Page 48270]]

The proposed amendment changes the surveillance Frequency from

18 months to 24 months for SRs in the Units 2 and 3 TS that are

normally a function of the refueling interval. In addition, the

proposed amendment changes the surveillance Frequency from 18 months

to 24 months for those SRs in the Unit 1 TS that control the test

interval for components and systems that are common to Units 1, 2,

and 3. Under certain circumstances SR 3.0.2 would allow a maximum

surveillance interval of 30 months for these SRs. The evaluations in

Section III [Licensee's August 14, 1998 application, Section III,

Safety Analysis] have shown that the reliability of protective

instrumentation will be preserved for the maximum allowable

surveillance interval. The proposed changes do not involve any

change to the design or functional requirements of plant systems,

and the surveillance test methods will be unchanged. The proposed

changes will not give rise to any increase in operating power level,

fuel operating limits, or effluents. In addition, the proposed

changes will not significantly increase any radiation levels. Based

on the foregoing considerations and the evaluations completed in

accordance with the guidance of Generic Letter 91-04, it is

concluded that the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed amendment requires no change to the plant design or

the mode of operation, for any item of equipment. The proposed

changes do not create the possibility of any new failure mechanisms.

Therefore, the proposed amendment does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

(3) The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed amendment seeks to change instrument calibration

surveillance intervals from 18 to 24 months. The primary

consideration relative to safety margin is that of exceeding

analytical limits for the current safety analyses as a result of

increased instrument drift over the extended surveillance interval.

The drift studies discussed in Section III.A have shown that the

existing setpoints and TS allowable values can be retained without

challenging the current analytical limits; thereby preserving the

assumptions in the current safety analyses and ensuring that safety

limits will not be exceeded.

To confirm that the drift errors remain within projected values,

instruments subjected to the longer interval between calibrations

will continue to be monitored as required by current plant

procedures. This practice will assure that no significant reduction

in safety margin is incurred by adoption of the proposed amendment.

Therefore, it is concluded that the proposed amendment does not

involve a significant reduction in a margin of safety.

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

its 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: Athens Public Library, 405 E.

South Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

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

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

Date of application for amendments: August 22,1996 (TS 97-04), as

supplemented on August 27, 1998.

Brief description of amendments: The amendments would change the

Sequoyah (SQN) Technical Specifications (TS) by extending the emergency

diesel generator allowed outage time from 72 hours to 7 days. This

amendment request was previously noticed on October 9, 1996 (61 FR

52969). The scope of the amendment request was changed by the August

27, 1998 submittal.

Basis for proposed no significant hazards consideration

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

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

of no significant hazards consideration, which is presented below:

TVA has concluded that operation of SQN Units 1 and 2, in

accordance with the proposed change to the TSs [Technical

Specifications] and operating licenses, does not involve a

significant hazards consideration. TVA's conclusion is based on its

evaluation, in accordance with 10 CFR 50.91(a)(1), of the three

standards set forth in 10 CFR 50.92(c).

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The EDGs [emergency diesel generators] supply backup power to

the essential safety systems in the event of a loss-of-offsite

(normal) power. The EDGs are not postulated to be an initiator of a

design basis accident. The requested change to provide a 7-day AOT

[allowed outage time] for the EDGs and the deletion of the

additional 72-hour extension for this AOT will not impact the plant

design, components or operational practices. The increased out-of-

service time does not invalidate assumptions used in evaluating the

radiological consequences of an accident and does not provide a new

or altered release path. In addition, the administrative changes to

delete EDG reporting requirements and an obsolete License Condition

will not impact plant equipment or operating practices. Therefore,

this change does not involve an increase in the probability of any

accident previously evaluated.

An increase in the AOT for the EDGs would not change the

conditions, operating configuration, or minimum amount of operable

equipment assumed in the plant Final Safety Analysis Report for

accident mitigation. The longer AOT would provide a longer time

window for maintenance, but would lesson the overall EDG

unavailability, therefore, it would reduce plant risk. The CDF [core

damage frequency] associated with a 7-day AOT increases from the

base case in the SQN [Sequoyah Nuclear Plant] IPE [individual plant

examination] but is not risk-significant. This CDF increase is based

on sensitivity studies performed in accordance with the guidance in

Draft Regulatory Guide DG-1065, dated June 1997. These studies

assume additional unavailability of the EDGs for an increase in AOT

even though plant practices are not expected to change. The EDG

availability improvements and CDF reductions during 12- and 6-year

maintenance activities compensates for this potential increase to

provide an overall safety benefit.

The deletion of the footnote for extending the AOT for fuel tank

cleaning removes inappropriate extensions of EDG out-of-service

time. SQN's implementation of the Maintenance Rule, 10 CFR 50.65,

also supports the proper scheduling and performance of maintenance

activities to ensure EDG unavailability is adequately controlled.

Based on no change in plant risk during routine maintenance, because

work activity durations are unchanged, and the decrease in overall

plant risk during the 12-and 6-year maintenance activities, as a

result of the 7-day EDG action time, this change will not result in

a significant increase in the consequences of an accident. In

addition, the administrative deletions of reporting requirements

that are not necessary based on Maintenance Rule implementation and

obsolete License Condition deletion will not increase the

consequences of an accident.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change to extend the AOT for the EDGs and delete

unnecessary TS and operating license provisions does not alter the

physical design or configuration of the plant. The EDG operation

remains unchanged, therefore, this change does not create the

possibility of a new or different kind of accident from any

previously analyzed.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed extension of the EDG action time for inoperable

units to 7 days will not alter plant equipment, setpoints or

operating practices that provide the necessary margin of safety. The

extension will reduce EDG unavailability and plant risk such that

the

[[Page 48271]]

EDG's ability to react to accident situations is increased. Overall

CDF, as a result of a 7-day AOT, indicates a slight increase but it

is not significant. The AOT extension deletion for fuel tank

cleaning is a conservative change to maintain appropriate EDG out-

of-service times. The deletions of administrative requirements for

reporting EDG reliability and obsolete License Conditions do not

impact functions that maintain the margins of safety and have been

or are continuing to be satisfied by other regulatory requirements.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon.

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

Callaway County, Missouri

Date of application request: June 29, 1998.

Description of amendment request: The amendment would revise

technical specification 3.7.1.7 to (1) address operability of all four

atmospheric steam dump (ASD) lines, (2) retain an action statement for

excessive ASD seat leakage, and (3) incorporate action statements for

multiple inoperable ASD lines.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Revising the LCO to refer to the ASD lines rather than the ASD

valves; requiring four ASD lines to be operable rather than three;

limiting the LCO 3.0.4 exception to one ASD line inoperable; and

adding a surveillance for the manual isolation valves constitutes a

more restrictive change from the current Specification. The proposed

changes impose more stringent requirements to ensure that ASD

operability is maintained consistent with the safety analysis and

licensing basis, and also to address all potential single failure

scenarios.

Therefore these changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

When two ASD lines are inoperable due to causes other than

excessive ASD seat leakage, the proposed change increases the

allowed outage time for restoration of all but one required ASD line

from 24 hours to 72 hours. The increase in time is not significant

when balanced against the availability of the condenser steam dump

system and/or the main steam safety valves, and the low probability

of an event occurring during the restoration period that would

require the ASD lines. Therefore the increase in allowed outage time

for restoration of all but one ASD line does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change revising the required completion time from

hot standby to hot shutdown from six hours to twelve hours is

consistent with NUREG-1431, Rev. 1, where the required completion

time to shut the plant down is revised to achieving hot standby in

six hours and hot shutdown within the following twelve hours. The

proposed change does not alter the plant configuration or operation

or the function of any safety system. Consequently, the change does

not increase the probability of an accident as defined in the

accident analysis. The proposed change permits a longer time to

cooldown to RHR entry conditions; however, this would not affect the

consequences of any postulated accidents and is appropriate due to

the need to avoid any transients while cooling down. Therefore the

proposed change would not involve a significant increase in the

probability or consequences of an accident.

Therefore, it is concluded that all of the above-proposed

changes do not significantly increase the probability or

consequences of any 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.

Revising the LCO to refer to the ASD lines rather than the ASD

valves; requiring four ASD lines to be operable rather than three;

limiting the LCO 3.0.4 exception to one ASD line inoperable; and

adding a surveillance for the manual isolation valves does not

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

of equipment will be installed) or changes in controlling

parameters. The proposed change does impose different requirements.

However, these changes are consistent with assumptions made in the

safety analysis and licensing basis. Thus, this change does not

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

any accident previously evaluated.

When two ASD lines are inoperable due to causes other than

excessive ASD seat leakage, the proposed change increases the

allowed outage time for restoration of all but one required ASD line

from 24 hours to 72 hours. The increase in time is not significant

when balanced against the availability of the condenser steam dump

system and/or the main steam safety valves, and the low probability

of an event occurring during the restoration period that would

require the ASD lines. The increase in the allowed outage time does

not result in a condition not previously considered or analyzed, and

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

of accident.

The proposed change revising the required completion time from

hot standby to hot shutdown from six hours to twelve hours is

consistent with NUREG-1431, Rev. 1, where the required completion

time to shut the plant down is revised to achieving hot standby in

six hours and hot shutdown within the following twelve hours. The

proposed change does not require physical alteration to any plant

system or change the method by which any safety-related system

performs its function. The change does allow additional time to

complete the transfer from the steam generator method for heat

removal to the RHR system, but does not alter the basic methodology.

Therefore, the proposed change would not create the possibility of a

new or different kind of accident.

All of the proposed changes discussed above do not create the

potential for a new or previously unanalyzed accident.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Revising the LCO to refer to the ASD lines rather than the ASD

valves; requiring four ASD lines to be operable rather than three;

limiting the LCO 3.0.4 exception to one ASD line inoperable; and

adding a surveillance for the manual isolation valves imposes more

stringent requirements. These requirements either have no impact on

or increase the margin of safety by increasing the scope of the

specification to include additional plant equipment; by adding

additional requirements; and by imposing a new surveillance. The

change is consistent with the safety analysis and licensing basis,

and does not involve a reduction in a margin of safety.

When two ASD lines are inoperable due to causes other than

excessive seat leakage, the proposed change increases the allowed

outage time for restoration from 24 hours to 72 hours. The increase

in time is not significant when balanced against the availability of

the condenser steam dump system and/or the main steam safety valves,

and the low probability of an event occurring during the restoration

period that would require the ASD lines. The increase in the allowed

outage time does not result in a condition not previously considered

and does not involve a significant reduction in a margin of safety.

The proposed change revising the required completion time from

hot standby to hot shutdown from six hours to twelve hours is

consistent with NUREG-1431, Rev. 1, where the required completion

time to shut the plant down is revised to achieving hot standby in

six hours and hot shutdown within the following twelve hours. The

change does not alter the basic regulatory requirements or change

any accident analysis assumptions, initial conditions or results.

Therefore, the proposed change would have no significant adverse

effect on margins of safety.

None of the proposed changes have any significant adverse effect

on margins of safety.

[[Page 48272]]

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Elmer Ellis Library,

University of Missouri, Columbia Missouri 65201.

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

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

NRC Project Director: William H. Bateman.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: July 28, 1998.

Description of amendment request: The North Anna Power Station

(NAPS), Unit 1 and 2, Technical Specifications (TS) Surveillance

Requirement (SR) 4.6.2.2.1.b requires verification, during

recirculation flow, that each outside recirculation spray (ORS) pump

develops a discharge pressure of greater than or equal to 115 pounds

per square inch (psig) and that each Casing Cooling pump develops a

discharge pressure of greater than or equal to 58 psig for Unit 1 and

46 psig for Unit 2 when tested. The proposed changes will revise the

testing acceptance criteria being verified from discharge pressure to

the required developed head. The frequency of testing shall be in

accordance with the Inservice 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:

Virginia Electric and Power Company has reviewed the

requirements of 10 CFR 50.92 as they relate to the proposed changes

for the North Anna Units 1 and 2 and determined that the changes do

not pose a significant hazards consideration * * * Specifically,

operation of the North Anna Power Station in accordance with the

proposed Technical Specification changes will not:

(a) Involve a significant increase in the probability or

consequences of an accident as previously evaluated

The applicable UFSAR [Updated Final Safety Analysis Report]

accidents previously evaluated are the LOCA [loss-of-coolant

accident] and MSLB [main steamline break]. The proposed changes

ensure that the Casing Cooling and ORS pumps will perform properly

with no unacceptable degradation by using the correct pump test

acceptance criteria as controlled by the PT program. This does not

increase the probability of a LOCA or MSLB.

(b) Create the possibility of a new or different type from any

accident previously evaluated

The proposed changes to the Technical Specifications will ensure

that the Casing Cooling and ORS pumps are tested at the frequency

established by the lnservice Testing Program to confirm their

ability to provide design basis flow during a LOCA/MSLB. This will

not result in any physical alteration to any plant system, nor would

there be a change in the method by which any safety related system

performs its function. The design and operation of the Casing

Cooling and ORS systems are not being changed. Also, the proposed

changes do not affect the design, operation or failure modes of the

Casing Cooling and ORS pumps and other components within the Casing

Cooling and ORS systems. Therefore, the proposed changes do not

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

any accident previously evaluated.

(c) Involve a significant reduction in a margin of safety

Implementation of the proposed changes ensures that the Casing

Cooling and ORS pumps do not operate with unacceptable degraded

flows during a LOCA/MSLB that are less than their containment

analysis design basis flow. Therefore, the proposed changes would

not 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 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: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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

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

Virginia 23219.

NRC Project Director: Pao-Tsin Kuo, Acting.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: July 28, 1998.

Description of amendment request: The North Anna Power Station

(NAPS), Unit 1 and 2, Technical Specifications (TS) Surveillance

Requirements (SR) 4.8.1.1.2.a.4, 4.8.1.1.2.c, 4.8.1.1.2.d.2,

4.8.1.1.2.d.4.b, 4.8.1.1.2.d.5, 4.8.1.1.2.d.6.b, 4.8.1.1.2.d.11.b, and

4.8.1.1.2.e currently require each Emergency Diesel Generator (EDG) to

be demonstrated OPERABLE by the performance of specific Surveillance

Requirements. One significant part of demonstrating operability of the

EDG requires verification that the frequency is within a specified

range, which is currently 60 plus or minus 1.2 Hz. The proposed changes

would change the frequency limit from 60 plus or minus 1.2 Hz to 60

plus or minus 0.5 Hz and separate the requirement of the EDG start from

the steady state voltage and frequency 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:

Virginia Electric and Power Company has reviewed the proposed

Technical Specification changes against the requirements of 10 CFR

50.92 and has determined that the proposed changes would not pose a

significant hazards consideration. Specifically, operation of the

North Anna Power Station in accordance with the proposed Technical

Specifications changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change provides a more stringent requirement for

the EDG frequency limit at steady state operation of 60 [plus or

minus] 0.5 Hz from the current 60 [plus or minus] 1.2 Hz. The change

additionally provides a separation of the start requirements from

the steady state limits for voltage and frequency. The change to the

EDG frequency limit does not result in operation that will increase

the probability of initiating an analyzed event and does not alter

assumptions relative to mitigation of an accident or transient

event. The change to the frequency limit is acceptable because the

safety analyses assumptions for emergency power limits the frequency

variations to 60 [plus or minus] 0.5 Hz and assumes that the EDG

supplies the emergency bus with electrical power within 10 seconds

of receiving an emergency start signal. The EDG output breaker will

close with no electrical power applied to the emergency bus when the

EDG output reaches 95% of rated voltage. The minimum frequency

requirement of 59.5 Hz is based on the steady state limit for the

EDG. The EDG supplies the electrical power for the required

equipment to mitigate the consequences of design basis events. The

minimum voltage and frequency (3740 volts and 59.5 Hz) limits ensure

that the ESF [engineered safety feature] equipment is maintained

with the required electrical power to mitigate the consequences of

an accident previously evaluated. Therefore, this change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. Create the possibility of a new or different type from any

accident previously evaluated.

[[Page 48273]]

The proposed change provides a more stringent requirement for

the EDG frequency at steady state operation of 60 [plus or minus]

0.5 Hz from the current 60 [plus or minus] 1.2 Hz. The change

additionally provides a separation of the start requirements from

the steady state limits for voltage and frequency. The change does

not introduce a new mode of plant operation and does not involve

physical modification to the plant. The proposed change does impose

different requirements. However, these changes are consistent with

the assumptions in the safety analyses. Thus, this change does not

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

any accident previously evaluated.

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

The proposed change provides a more stringent requirement for

the EDG frequency at steady state operation of 60 [plus or minus]

0.5 Hz from the current 60 [plus or minus] 1.2 Hz. The change

additionally provides a separation of the start requirements from

the steady state limits for voltage and frequency. The change to the

frequency limit is acceptable because the safety analyses

assumptions for emergency power limits the frequency variations to

60 [plus or minus] 0.5 Hz and assumes that the EDG supplies the

emergency bus with electrical power within 10 seconds of receiving

an emergency start signal. The EDG output breaker will close with no

electrical power applied to the emergency bus when the EDG output

reaches 95% of rated voltage. The minimum frequency requirement of

59.5 Hz is based on the steady state limit for the EDG.

The EDG supplies the electrical power for the required equipment

to mitigate the consequences of design basis events. The minimum

voltage and frequency (3740 volts and 59.5 Hz) limits ensure that

the ESF equipment will be supplied with the required electrical

power to mitigate previously evaluated accidents. The margin of

safety is established through the design of the plant structures,

systems and components, the parameters within which the plant is

operated, and the establishment of the setpoints for the actuation

of equipment relied upon to respond to an event. The change allowing

the separation of the start requirements from the steady state

voltage and frequency limits, due to the short time period allowed

in this condition, does not significantly impact the performance of

structures; systems or components relied upon for accident

mitigation or any safety analysis assumptions. Therefore, the change

does not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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

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

Virginia 23219.

NRC Project Director: Pao-Tsin Kuo, Acting.

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.

Duke Energy Corporation, Docket Nos. 50-413 and 50-414, Catawba Nuclear

Station, Units 1 and 2, York County, South Carolina

Date of amendment request: August 6, 1998.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) Surveillance Requirement

4.8.1.1.2.i.2. This requirement is in conflict with a relief granted by

the NRC staff in February 1995. The deletion of TS Surveillance

Requirement 4.8.1.1.2.i.2 would remove such a conflict.

Date of publication of individual notice in Federal Register:

August 17, 1998 (63 FR 43962).

Expiration date of individual notice: September 16, 1998.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego County, New York

Date of application for amendment: July 16, 1998. This notice

supersedes a previous notice (62 FR 40851, published July 30, 1997)

that was based upon an amendment request dated July 2, 1997. The

request dated July 2, 1997, was superseded in its entirety by the

amendment request dated July 16, 1998.

Brief description of amendment: The amendment would change

Technical Specification 3/4.2.3 regarding reactor coolant chemistry in

accordance with a report by Electrical Power Research Institute, Inc.

TR-103515-R1, ``BWR Water Chemistry Guidelines, 1996 Revision,'' also

known as Boiling Water Reactor Vessel and Internals Project-29.

Date of publication of individual notice in Federal Register:

August 13, 1998 (63 FR 43432).

Expiration date of individual notice: September 14, 1998.

Local Public Document Room location: Reference and Documents

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

York 13126.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: February 27, 1998, as supplemented July

14, 1998.

Brief description of amendment requests: The proposed amendments

would allow a design modification to the existing Anticipated Transient

Without Scram (ATWS) Mitigation System Actuation Circuitry (AMSAC). The

design modification would install a Diverse Scram System (DSS) designed

to meet the requirements of a DSS described by 10 CFR 50.62 (ATWS Rule)

for non-Westinghouse designed plants and make major modifications to

the existing AMSAC.

Date of publication of individual notice in Federal Register:

August 17, 1998 (63 FR 4365).

Expiration date of individual notice: September 16, 1998.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: October 14, 1997, as

supplemented July 23, 1998.

Brief description of amendment: The proposed amendment would change

the James A. FitzPatrick Technical Specifications to provide for

installation of additional racks to increase spent fuel pool capacity,

and to correct the maximum exposure dependent, infinite lattice

multiplication factor for fuel bundles.

[[Page 48274]]

Date of initial notice in Federal Register: August 24, 1998 (63 FR

45096).

Expiration date of individual notice: September 23, 1998.

Local Public Document Room location: Reference and Documents

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

York 13126.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

Date of application for amendments: January 14, 1998.

Brief description of amendments: The amendments revise the

Technical Specifications to support replacement of the 125 volt direct

current (Vdc) AT&T batteries with new Charter Power Systems, Inc. (C&D)

batteries. In addition, the crosstie loading limitation is revised to

reflect the larger capacity of the C&D batteries.

Date of issuance: August 18, 1998.

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

Amendment Nos.: 94 and 94.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: May 20, 1998 (63 FR

27758). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 18, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wilmington Public Library, 201

S. Kankakee Street, Wilmington, Illinois 60481.

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

Michigan

Date of application for amendment: April 2, 1998 (NRC-98-0057).

Brief description of amendment: The amendment revises Technical

Specification 3.3.7.5 to permit entering Operational Conditions 1 and 2

prior to completion of Surveillance Requirements for the primary

containment hydrogen and oxygen monitors in order to establish the

conditions necessary (inerted containment) to properly perform the

calibrations. The amendment also allows an increase in the frequency of

the calibration for the oxygen monitors from once every 18 months to

quarterly and corrects the nomenclature for the hydrogen and oxygen

monitors in tables 3.3.7.5-1 and 4.3.7.5-1.

Date of issuance: August 20, 1998.

Effective date: August 20, 1998, with full implementation within 90

days.

Amendment No.: 125.

Facility Operating License No. NPF-43. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19968).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 20, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

Ellis Reference and Information Center, 3700 South Custer Road, Monroe,

Michigan 48161.

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

Michigan

Date of application for amendment: March 27, 1998 (NRC-98-0034), as

supplemented May 28 and July 31, 1998.

Brief description of amendment: The amendment revises footnotes

associated with the emergency core cooling system (ECCS) in Technical

Specifications 3.5.1, ``ECCS--Operating,'' and 3.5.2, ``ECCS--

Shutdown,'' to indicate that a low pressure coolant injection system

loop may be considered operable during alignment and operation for

decay heat removal if it is capable of being manually realigned and is

not otherwise inoperable. The associated Bases are also revised.

Date of issuance: August 25, 1998.

Effective date: August 25, 1998, with full implementation within 90

days.

Amendment No.: 126.

Facility Operating License No. NPF-43: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19968). The May 28 and July 31, 1998, letters provided clarifying

information that was within the scope of the original Federal Register

notice and did not change the staff's initial proposed no significant

hazards considerations determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 25, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

Ellis Reference and Information Center, 3700 South Custer Road, Monroe,

Michigan 48161.

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

Michigan

Date of application for amendment: September 25, 1996 (NRC-96-

0085), as supplemented by letters dated November 26, 1997, and March 10

and June 17, 1998.

Brief description of amendment: The amendment revises Surveillance

Requirement 4.8.4.3 to clarify the situational testing requirement for

thermal overload devices to indicate that this portion of the

requirement must be completed upon initial installation of a thermal

overload device and following any maintenance that could affect its

performance.

NRC has also granted the request of Detroit Edison Company to

withdraw a portion of its September 25, 1996,

[[Page 48275]]

application. The proposed change would have deleted the requirement for

periodically testing motor-operated valve thermal overload protective

devices. However, by letter dated June 17, 1998, the licensee withdrew

this portion of the amendment request. For further details with respect

to these actions, see the application for amendment dated September 25,

1996, as supplemented above, and the licensee's letter dated June 17,

1998, which withdrew this portion of the application for license

amendment, and the staff's safety evaluation enclosed with the

amendment. The above documents 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 listed

below.

Date of issuance: August 25, 1998.

Effective date: August 25, 1998, with full implementation within 90

days.

Amendment No.: 127.

Facility Operating License No. NPF-43: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55030).

The November 26, 1997, and March 10 and June 17, 1998, submittals

provided additional clarifying information within the scope of the

original Federal Register notice and did not change the staff's initial

proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 25, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

Ellis Reference and Information Center, 3700 South Custer Road, Monroe,

Michigan 48161.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: April 8, 1998.

Brief description of amendments: The amendments revise Technical

Specification Section 3/4.6.5.1, regarding the ice condenser, to reduce

the total ice weight from 2,475,252 to 2,330,856 pounds, and to reduce

individual ice basket ice weight from 1273 to 1199 pounds. The

associated Bases section is also revised to reflect the changed

requirements.

Date of issuance: August 25, 1998.

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

30 days from the date of issuance.

Amendment Nos.: Unit 1--168; Unit 2--160.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR

25107).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 25, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: December 11, 1997.

Brief description of amendments: The amendments revise Technical

Specification Table 3.3-4, Engineered Safety Feature Actuation System

Instrumentation Trip Setpoints, to require that suction of the Nuclear

Service Water System be swapped from Lake Wylie to the Standby Nuclear

Service Water Pond at a higher minimum water level of Lake Wylie.

Specifically, the amendments change the swap setpoint from greater than

or equal to 554.4 feet to greater than or equal to 557.5 feet, and the

allowable value from greater than or equal to 552.9 feet to greater

than or equal to 555.4 feet.

Date of issuance: August 25, 1998.

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

30 days from the date of issuance.

Amendment Nos.: Unit 1--169; Unit 2--161.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 11, 1998 (63

FR 6983).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 25, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: September 15, 1997, as

supplemented by letters dated March 5, April 27, June 15, July 22, and

August 10, 1998.

Brief description of amendments: The amendments revise Technical

Specification Figures 3.4-2 and 3.4-3 (pressure-temperature limits

curves), Table 4.4-5 (reactor vessel surveillance capsule withdrawal

schedule), and Sections 3/4.4.9.3 and 3.5.3 (requirements concerning

overpressure protection). The associated Bases are also revised.

Date of issuance: August 28, 1998.

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

60 days.

Amendment Nos.: Unit 1--170; Unit 2--162.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 8, 1997 (62 FR

52580); and July 29, 1998 (63 FR 40553).

The March 5, April 27, July 22, and August 10, 1998, letters

provided additional information that did not change the scope of the

September 15, 1997, application and the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 28, 1998.

No significant hazards consideration commets received: No.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Duke Energy Corporation, Docket No. 50-287, Oconee Nuclear Station,

Unit 3, Oconee County, South Carolina

Date of application of amendment: July 20, 1998.

Brief description of amendment: The amendment extends, on a one-

time basis, Technical Specification Surveillance 4.18.3 for hydraulic

and mechanical snubber testing. The tests are required to be performed

at a frequency of 18 months, with a maximum allowed frequency of 22

months, 15 days. The amendment extends this to a maximum of 25 months.

Date of Issuance: August 26, 1998.

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

30 days from the date of issuance.

Amendment No.: 229.

[[Page 48276]]

Facility Operating License No. DPR-55: The amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: July 27, 1998 (63 FR

40137).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 26, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Duke Energy Corporation, Docket No. 50-287, Oconee Nuclear Station,

Unit 3, Oconee County, South Carolina

Date of application of amendment: July 16, 1998.

Brief description of amendment: The amendment extends, on a one-

time basis, during Operating Cycle 17, certain specified Technical

Specification surveillances that are required to be performed at a

frequency of 18 months from the maximum allowed frequency of 22 months,

15 days, to a maximum of 24 months.

Date of Issuance: August 28, 1998.

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

30 days from the date of issuance.

Amendment No.: 230.

Facility Operating License No. DPR-55: The amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: July 29, 1998 (63 FR

40555)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

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

Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2),

Shippingport, Pennsylvania

Date of application for amendments: December 19, 1997, as

supplemented June 16, July 9, and July 15, 1998.

Brief description of amendments: These amendments revise the

requirements for the source range neutron flux channels in Modes 2

(Below P-6), 3, 4, and 5 to incorporate the guidance provided in NUREG-

1431, the NRC's improved Standard Technical Specifications with some

modifications to address plant-specific design features. This change

allows (1) the use of alternate detectors provided the required

functions are provided, and (2) plant cooldown with inoperable

detectors provided the shutdown margin accounts for the temperature

change. This change also modifies the BVPS-2 Technical Specification

(TS) Table 3.3-1 Channels To Trip and Minimum Channels Operable

requirements to 0 and 1, respectively. This portion of the amendment

makes these BVPS-2 requirements consistent with the current BVPS-1

requirements. For both BVPS-1 and BVPS-2, TS Table 4.3-1 is modified to

include a notation exempting the alternate source range detectors from

surveillance testing until they are required for operability.

Date of issuance: August 26, 1998.

Effective date: Both units, effective immediately, to be

implemented within 60 days.

Amendment Nos.: 217 and 94.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: March 11, 1998 (63 FR

11918).

The June 16, July 9, and July 15, 1998, letters provided clarifying

information that did not change the initial no significant hazards

consideration determination or expand the amendment request beyond the

scope of the March 11, 1998, Federal Register notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 26, 1998.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

Date of application for amendments: June 3, 1998.

Brief description of amendments: Revise the surveillance

requirements of TS Section 4.11.2.5.1, Explosive Gas Mixture, to add a

reference the St. Lucie Units 1 and 2 Updated Final Safety Analysis

Reports for clarification of an alternative monitoring method to be

used in the event that continuous monitoring of explosive gas mixtures

in the waste decay tanks becomes inoperable.

Date of Issuance: August 10, 1998.

Effective Date: August 10, 1998, and shall be implemented within 30

days of receipt.

Amendment Nos.: 156 and 94.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 1, 1998 (63 FR

35990).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 10, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

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

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie

Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendment: March 3, 1998.

Brief description of amendment: This amendment revises the TS in

three areas. First, the amendment revises TS 3.4.7, Reactor Coolant

System-Chemistry, to eliminate the need for sampling of reactor coolant

system chemistry in the defueled condition. Second, the amendment

revises TS 5.6.1.a.1, Design Features-Fuel Storage-Criticality, to

reflect the total uncertainty associated with the unborated criticality

analysis previously approved by NRC. And third, the amendment revises

TS 6.5.2.9.d, Technical Review Responsibilities, to be consistent with

the quality assurance proce

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