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. Pub. L. 97-415 revised section 189 of the Atomic

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from November 20, 1998, through December 4, 1998.

The last biweekly notice was published on December 2, 1998 (63 FR

66590).

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 January 15, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

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Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,

Maryland.

Date of amendments request: November 19, 1998.

Description of amendments request: The proposed amendment revises

Technical Specification 3.7.6, ``Service Water (SRW) System'' to allow

operation of Calvert Cliffs with one SRW plate and frame heat exchanger

(PHE) secured for maintenance or other reasons, and removing one

containment air cooler (CAC) from service to enable the affected

subsystem to remain operable. Specifically, the proposed change adds

``One SRW heat exchanger inoperable'' as a new condition for Limiting

Condition for Operation (LCO) 3.7.6. The required actions for the new

condition are to secure one CAC within one hour and restore the heat

exchanger to operable condition within 7 days, or be in Mode 3 in 6

hours and Mode 5 in 36 hours. This limits the effect of one inoperable

PHE to only one containment cooling train made inoperable by the PHE.

Consequently, the new action statement introduced in the SRW LCO for an

inoperable PHE is similar to the one that already exists in the CAC LCO

for one inoperable containment cooling train.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

None of the systems associated with the proposed revision to the

Calvert Cliffs Technical Specifications are accident initiators. The

Saltwater (SW) and SRW systems are used to mitigate the effects of

accidents analyzed in the Updated Final Safety Analysis Report

(UFSAR). The SW and SRW Systems provide cooling to safety-related

equipment following an accident. The CACs are provided with SRW to

remove heat from the Containment in the event of an accident. They

support accident mitigation functions; therefore, the proposed

modification does not increase the probability of an accident

previously evaluated.

The proposed revision will provide greater availability of

safety-related equipment during PHE maintenance activities. It

ensures that the safety features provided by the SW and SRW, except

for the isolated CAC, are maintained, i.e., the availability of

safety-related equipment required to mitigate the radiological

consequences of an accident described in the UFSAR is enhanced by

the flexibility provided by this Technical Specification revision.

Furthermore, the proposed revision will not change, degrade, or

prevent actions described or assumed in any accident described in

the UFSAR. The proposed activity will not alter any assumptions

previously made in evaluating the radiological consequences of any

accident described in the UFSAR.

Therefore, the proposed modification does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

None of the systems associated with this modification are

identified as accident initiators in the UFSAR. The SW and SRW

Systems and the CACs are used to mitigate the effects of accidents

analyzed in the UFSAR. None of these functions required of these

systems have been changed by the proposed revision to the Technical

Specifications. This activity does not modify any system, structure,

or component such that it could become accident initiator, as

opposed to its current role as an accident mitigator.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

safety.

The safety design basis for the SW and SRW Systems is the

availability of sufficient cooling capacity to ensure continued

operation of equipment during normal and accident conditions. The

redundant cooling capacity of these systems, assuming a single

failure, is consistent with assumptions used in the accident

analysis.

With one SRW subsystem inoperable, the remaining SRW subsystem

is adequate to perform the heat removal function. However, the

reliability is reduced because a single failure in the operable SRW

subsystem could result in loss of SRW function. The proposed change

will allow continued operation of some SRW-cooled components while a

PHE is being out-of-service. The second SRW subsystem will still be

available to perform the SRW function. In addition, the reliability

of many diesel generator-backed components will be improved since

the second diesel generator will remain operable while in this

action statement.

During a design basis accident, a minimum of one containment

cooling train (two of the four CACs) and one containment spray

train, is required to maintain the containment peak pressure and

temperature, below the design limits. Under the existing Technical

Specification requirement, with one containment cooling train

inoperable, the inoperable containment cooling train must be

returned to operable status within seven days. The remaining

operable containment spray and cooling units provide iodine removal

capabilities and are capable of removing at least 100% of the heat

removal needs after an accident. The seven-day completion time was

developed taking into account the redundant heat removal

capabilities afforded by combinations of the containment spray and

cooling systems, and the low probability of a design basis accident

occurring during this period. The proposed change to Technical

Specification 3.7.6 would allow three CACs to remain operable during

maintenance on a PHE, instead of the two that are maintained under

the current Technical Specification requirement.

[[Page 69334]]

For the above reasons, the margin of safety has been preserved,

and in some cases increased, by the proposed revision to the

Technical Specifications.

Therefore, this proposed modification does not significantly

reduce the margin of safety.

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

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

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: S. Singh Bajwa, Director.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,

Maryland.

Date of amendments request: November 20, 1998.

Description of amendments request: On September 9, 1996, a final

rule amending 10 CFR 50.55a was issued requiring owners to implement,

by September 9, 2001, the requirements of the 1992 Addenda of the

American Society of Mechanical Engineers Boiler and Pressure Vessel

Code Section XI, Subsections IWE and IWL, as modified and supplemented

by 10 CFR 50.55a. Baltimore Gas and Electric Company (BGE) have

developed a program plan to effect the implementation of Subsection IWE

and IWL. BGE's submittal requests a license amendment in support of the

program plan. One Technical Specification (TS) change requested is an

administrative change that removes a TS originally developed from

Regulatory Guide (RG) 1.35. Compliance with RG 1.35 is not sufficient

to comply with 10 CFR 50.55a, as amended. The other TS changes request

the removal from the TSs requirements that are a duplication of 10 CFR

50.55a.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The Containment Building is a passive safety structure that

prevents the release of radioactive materials to the environment in

post-accident conditions. The proposed Technical Specification

changes delete requirements of the Technical Specifications that

have been made obsolete by the improvements of the Containment

Building inspections required by the changes in the regulations. The

improved inspections required by the American Society of Mechanical

Engineers Code serve to maintain Containment response to accident

conditions, by causing the identification and repair of defects in

the Containment Buildings.

Relocating existing requirements, eliminating requirements that

duplicate regulations, and making administrative improvements

provide Technical Specifications that are easier to use. Because

existing requirements are controlled by regulation, there is no

reduction in commitment and adequate control is still maintained.

Likewise, the elimination of requirements that duplicate regulations

enhances the usability of the Technical Specifications without

reducing commitments. Therefore, the proposed changes would not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

The Containment Building is a passive safety structure designed

to contain radioactive materials released from the Reactor Coolant

System. The performance of the Containment Building is not evaluated

as the causal factor in any accident at Calvert Cliffs Nuclear Power

Plant. The proposed Technical Specification changes delete

requirements of the Technical Specifications that have been made

obsolete by the improvements of the Containment Building inspections

required by the changes in the regulations. Revising the Technical

Specifications, to comply with current regulations and to eliminate

duplication of requirements, does not create the possibility of a

new or different type of accident from any accident previously

evaluated.

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

safety.

The safety function of the Containment Building is to provide a

boundary to the release of radioactive material to the environment

during post-accident conditions. The changes to the Technical

Specifications incorporate improved inspection techniques and

criterial to ensure optimum Containment integrity and, therefore,

optimum containment response in the event of an accident resulting

in a release of radioactive material from the Reactor Coolant

System.

Optimizing containment integrity will result in maintaining the

margin of safety allowed by the Containment Buildings. Therefore,

the proposed changes 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 50.92(c) are

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: S. Singh Bajwa, Director.

CBS Corporation acting through its Westinghouse Electric Company

Division (licensee), Westinghouse Test Reactor, Waltz Mill Site,

Westmoreland, Pennsylvania, Docket No. 50-22, License No. TR-2.

Date of amendment request: September 28, 1998, supplemented on

November 17, 1998.

Description of amendment request: CBS Corporation acting through

its Westinghouse Electric Company Division is the licensee for the

Westinghouse Test Reactor (WTR) at Waltz Mill, Pennsylvania. The

licensee is authorized to only possess the reactor and a

decommissioning plan has been approved. The licensee is planning to

sell most of its nuclear related facilities to other entities, but will

retain the WTR. One of the arrangements made with the purchasers of the

other facilities is that the Westinghouse name will be conveyed with

these facilities, and because of this arrangement, the licensee

requests that the license associated with the Westinghouse Test Reactor

be changed to simply CBS Corporation, to eliminate any reference to the

name Westinghouse.

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

considerations. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). A proposed amendment to a

license of a facility involves no significant hazards consideration if

operation of the facility in accordance with the proposed amendment

would not: (1) involve a significant increase in the probability or

consequences of an accident previously evaluated; or (2) create the

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

previously evaluated; or (3) involve a significant reduction in a

margin of safety.

The staff agrees with the licensee's no significant hazards

consideration determination submitted on November 17, 1998, for the

following reason.

This corporate name change does not involve any change in the

management, organization, location, facilities equipment, or procedures

related to the licensed activities under the WTR

[[Page 69335]]

license. The employees responsible for the licensed WTR facility will

still be responsible, either directly through the CBS Corporation or

through contractual arrangements for which CBS Corporation is

ultimately responsible, notwithstanding the new name of the licensee.

Based on a review of the licensee's analysis, and on the staff's

analysis detailed above, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Attorney for license: Lisa A. Campagna, Assistant General Counsel,

Law Department, CBS Corporation, P.O. Box 355, Pittsburgh, Pennsylvania

15230.

NRC Project Director: Seymour H. Weiss.

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

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

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

1 and 2, Will County, Illinois.

Date of amendment request: October 30, 1998.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TS) to reduce the spent fuel pool

(SFP) inadvertent draindown level to account for the effects of

potential failures of the SFP cooling and skimmer loops.

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?

This change to the TS does not involve an increase in the

probability of an accident previously evaluated. The initial

conditions of the limiting dewatering incidents involve initiating

circumstances/failures such as accidental gate openings, gate seal

failures, or an open transfer tube.

Specifying a revised inadvertent drain limit which meets the SRP

[Standard Review Plan, NUREG-0800] acceptance criteria is unrelated

to the probability of occurrence of the precursors or initiating

events. These initiators are not affected by the SFP cooling or

skimmer loop piping/component failure scenarios. There is no change

being made to the approved design, nor is there any operational

change being made which would increase the probability of

occurrence.

This change to the TS does not involve an increase in the

consequences of an accident previously evaluated. As documented in

NUREG-0876, Byron SER, Section 9.1.3, page 9-5, the anti-siphon

protection design of the SFP cooling and clean-up piping was

reviewed and found to be acceptable stating that ``all connections

to the spent-fuel pool are either near the normal water level or are

provided with antisiphon holes to preclude possible siphon draining

of the pool water.'' This review is applicable to Braidwood as

documented in NUREG-1002, Braidwood SER. The anti-siphon attributes

employed in the SFP skimmer loops at Braidwood, (under consideration

at Byron), are similar in design as well as their submergence levels

previously evaluated for the SFP cooling loops. The proposed change

revises the SFP inadvertent drain limit from approximately 423 feet

to 410 feet to bound the failure effects of both the SFP cooling and

skimmer loops, while considering any maloperation or failure

scenario. The revised value meets the SRP acceptance criteria of

maintaining at least 10 feet above the active fuel ensuring that

adequate radiation shielding is maintained as previously analyzed.

There is no physical or operational change being made which would

alter the sequence of events, plant response, or conclusions of the

affected analysis. There is no change in the type or amount of any

effluents released, and no change in either the Onsite or Offsite

dose consequences as a result of this change.

Therefore, based on this evaluation, this proposed amendment

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

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

different kind of accident from any accident previously evaluated.

This change specifically identifies the SFP level sufficient to

ensure that the SRP acceptance criteria for inadvertent draining are

met while accounting for the failure effects of both the SFP cooling

and skimmer loops. Any inadvertent SFP draining due to potential

failures of the SFP skimmer loops is similar in nature to the

inadvertent SFP draining effects previously considered due to

failures of the SFP cooling loops. No new equipment is being

installed, and no installed equipment is being operated in a new or

different manner with this change. There is no change in plant

operation that affects previously evaluated failure modes. This

change does not represent a new failure mode or accident from what

has been previously evaluated.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

of safety?

The current TS value does not address inadvertent SFP draining

due to potential failures of the SFP skimmer loops or cooling

suction lines as was done for the SFP cooling discharge lines. This

change specifically identifies the SFP level sufficient to ensure

that the SRP acceptance criteria for inadvertent draining are met

while accounting for the failure effects of both the SFP cooling and

skimmer loops in determining the proposed TS value. The most

limiting postulated SFP dewatering incidents involve SFP drainage to

either a dry transfer canal, a dry transfer canal and cask fill

area, or a dry transfer canal and cask fill area which additionally

communicates through an open transfer tube to an empty refuel

cavity. The initial conditions of the dewatering incident analysis

and resultant water levels over the spent fuel are not affected by

this SFP skimmer/cooling loop issue because these incident

initiators are not effected by the SFP cooling or skimmer loop

failures, thus preserving the previously analyzed and approved

margin for these dewatering incidents.

For the less-limiting SFP skimmer/cooling loop failure issue,

the proposed TS change inadvertent drain limit meets the SRP minimum

requirement of at least 10 feet above the top of the active fuel

ensuring that adequate radiation shielding is maintained. This

change would allow for the conservative acceptance criteria for the

current UFSAR [Updated Final Safety Analysis Report] design analysis

to continue to be met.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

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

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

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

Wilmington, Illinois 60481.

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

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

NRC Project Director: Stuart A. Richards.

Commonwealth Edison Company, Docket No. 50-374, LaSalle County

Station, Unit 2, LaSalle County, Illinois.

Date of amendment request: November 9, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specification 3/4.3.2, ``Isolation Actuation

Instrumentation'' to add/revise various isolation setpoints for leak

detection instrumentation. These changes are necessary due to

modifications to the Reactor Water Cleanup (RWCU) System to restore

``hot'' suction to the RWCU pumps and due to a re-evaluation of the

high energy line break analysis. In addition, the amendment would

eliminate isolation actuation trip functions for the Residual Heat

Removal (RHR) system steam

[[Page 69336]]

condensing mode and shutdown cooling mode.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

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

change in probability of an accident. A line break in the subject

areas, would consist of an instantaneous circumferential break

downstream of the outermost isolation valve of one of these systems.

The leak detection isolation is only a precursor of a break, and

thus does not affect the probability of a break.

(b) There is minimal effect on the consequences of analyzed

accidents due to changing the leak detection ambient temperature or

Delta T setpoint and allowable values to detect 25 gpm equivalent

leakage. The addition of more ambient temperature and T

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

break detection will actually decrease the consequences of the

associated accidents. The worst case accident outside the primary

containment boundary is a main steam line break which bounds the

dose consequences of all line breaks and therefore bounds any size

of leak.

The deletion of the RHR steam condensing mode isolation

actuation instrumentation trip functions from the LaSalle Technical

Specifications does not increase the probability or consequences of

an accident previously evaluated, because this mode of operation of

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

the lines that were previously high energy lines are isolated during

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

Operational Condition 2 (Startup mode), and Operational Condition 3

(Hot Shutdown).

The deletion of the RHR shutdown cooling mode leak detection T

and Delta T isolation actuation instrumentation trip functions from

the LaSalle Technical Specifications does not increase the

probability or consequences of an accident previously evaluated,

because the leak detection is only a precursor of a break, and thus

does not affect the probability of a break. Also, there are two

other methods of detecting abnormal leakage and isolating the system

in Technical Specification trip functions A.6.a, Reactor Vessel

Water Level--Low, Level 3 and A.6.c, RHR Pump Suction Flow--High. In

addition, other means to detect leakage from the RHR system, such as

sump monitoring and area radiation monitoring, are also available.

In accordance with Technical Specification Administrative

Requirement 6.2.F.1, LaSalle has a leakage reduction program to

reduce leakage from those portions of systems outside primary

containment that contain radioactive fluids. RHR, including piping

and components associated with the shutdown cooling mode, is part of

this program, which includes periodic visual inspection of the

system for leakage. The sump monitoring, radiation monitoring and

periodic inspections for system leakage makes the probability of a

leak of 5 gpm going undetected for more than a day very low.

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

which RHR shutdown cooling mode is able to operate, reactor coolant

makeup and outflow is very low compared to normal plant operation. A

change in flow balance due to a leak is thus more readily detectable

with reactor coolant water level changes and makeup flow rate, and

thus precludes a significant leak going undetected before break

detection instrumentation would cause automatic isolation.

Therefore, this proposed amendment 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 because:

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

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

detection and automatic isolation of the system as necessary in the

event of leakage in these areas. This change maintains this

capability with at least two different methods of detection of

abnormal leakage for protection from the flooding concerns of a

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

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

affected.

This change also maintains or adds primary containment isolation

logic for the leak detection isolation based on temperature

monitoring in RWCU areas and break detection based on RWCU pump

suction flow--high. The additional instrumentation and the

associated isolation logic is the same or similar to existing

instrumentation and logic for containment actuation instrumentation,

so no new failure modes are created in this way.

Therefore, these proposed changes do not create the possibility

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

evaluated.

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

because:

The change to the automatic isolation setpoint for high Delta T

leak detection in the heat exchanger rooms is based on current

configuration calculated/analyzed response to a small leak compared

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

heat exchanger rooms that is necessary to actuate isolation on

ambient temperature during winter conditions, does not adversely

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

the critical crack leakage rate as represented in UFSAR [Updated

Final Safety Analysis Report] Figure 5.2-11. Additionally,

differential temperature leak detection is conservative under these

same conditions, and will actuate isolation at a leakage rate less

than the established limit. The leak detection isolation logic is

unchanged and thus remains single failure proof.

The addition of automatic primary containment isolation on

ambient temperature and Delta T-High for the Reactor Water Cleanup

System (RWCU) Pump, Pump Valve, Holdup Pipe, and Filter/

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

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

with respect to leak detection and line breaks in the RWCU system,

because the system isolation diversity is increased and the amount

of system piping monitored for leakage is increased.

The setpoints for the ambient temperature and Delta T leak

detection isolations being changed or added and the RWCU pump

suction flow--high are set sufficiently high enough so as not to

increase the possibility of spurious actuation. In the event that a

spurious actuation does occur, little safety significance is

presented since the RWCU system performs no safety function. The

setpoints and allowable values for the proposed changes also assure

sufficient margin to the analytical values and are high enough to

prevent spurious actuations based on calculations consistent with

Regulatory Guide 1.105.

The deletion of the RHR steam condensing mode isolation

actuation instrumentation does not effect the margin of safety,

because this mode is no longer utilized by LaSalle in Operational

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

The elimination of the temperature based trip functions for the

RHR shutdown cooling mode area is based on the determination that

temperature is not the appropriate parameter for leak detection as

it does not provide meaningful indication and will not provide

setpoints that would be sufficiently above the normal range of

ambient conditions to avoid spurious isolations.

There are two other methods of detecting abnormal leakage and

isolating the system in Technical Specification trip function A.6,

which are A.6.a, Reactor Vessel Water Level--Low, Level 3 and A.6.c,

RHR Pump Suction Flow--High. In addition, other means to detect

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

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

Technical Specification Administrative Requirement 6.2.F.1, LaSalle

has a leakage reduction program to reduce leakage from those

portions of systems outside primary containment that contain

radioactive fluids. RHR, including piping and components associated

with the shutdown cooling mode, is part of this program, which

includes periodic visual inspection of the system for leakage.

The previous evaluation of diversity of isolation parameters, as

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

diversity of isolation parameters is maintained because there are at

least two different methods available to detect and allow isolation

of the system for a line break, as necessary.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

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

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

satisfied. Therefore, the NRC staff

[[Page 69337]]

proposes to determine that the requested amendment involves no

significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library, 815

North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,

Illinois 61348-9692.

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

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

NRC Project Director: Stuart A. Richards.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York.

Date of amendment request: October 9, 1998.

Description of amendment request: The proposed amendment would

revise Section 6.0, administrative controls, of the Technical

Specifications (TSs). Specifically, TS Sections 6.5.2.1.j, 6.7.1.c, and

6.8.1.a would be revised to correct typographical errors. In addition,

TS Section 6.5.2.2 would be revised to change the membership of the

Nuclear Facility Safety Committee (NFSC). This change would provide

Consolidated Edison (Con Ed) with the flexibility to obtain industry

experts outside of Con Ed to perform the duties of Chairman, or Vice

Chairman, and members of the NFSC.

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. There is no significant increase in the probability or

consequences of an accident previously evaluated.

The proposed amendment is administrative in nature. It involves

a change in 1) the Nuclear Facilities Safety Committee (NFSC)

Chairman or Vice Chairman to allow the services of an individual

other than a senior official of the Company, and 2) allowing NFSC

membership by other than Con Edison employees. In either case,

concurrence by the Senior Vice President, Nuclear Operations is

required.

These changes do not affect possible initiating events for

accidents previously evaluated or alter the configuration or

operating of the facility. The Limiting Safety Systems Settings and

Safety Limits specified in the current Technical Specifications

remain unchanged. Therefore, the proposed changes to the subject

Technical Specification would not increase the probability or

consequences of an accident previously evaluated.

2. The possibility of a new or different kind of accident from

any accident previously evaluated has not been created.

As stated above, the proposed changes are administrative in

nature. The safety analysis of the facility remains complete and

accurate. There are no physical changes to the facility, and the

plant conditions for which the design basis accidents have been

evaluated are still valid. The operating procedures and emergency

procedures are unaffected. Consequently, no new failure modes are

introduced as a result of the proposed changes. Therefore, the

proposed changes will not initiate any new or different kind of

accident.

3. There has been no significant reduction in the margin of

safety.

The proposed changes are administrative in nature. Since there

are no changes to the operation of the facility or physical design

the Updated Final Safety Analysis Report (UFSAR) design basis,

accident assumptions, or Technical Specification Bases are not

affected. Therefore, the proposed changes will not result in a

reduction in the margin of safety.

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

this review, it appears that the three standards of 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: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Project Director: S. Singh Bajwa, Director.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan.

Date of amendment request: November 9, 1998.

Description of amendment request: The proposed amendment would

delete the Chemical and Volume Control System (CVCS) operability

requirements currently in technical specifications (TS) 3.2 and 3.17.6,

and the associated surveillance testing requirements currently in TS

4.2 and 4.17. The requirements have been added to the Palisades

Operating Requirements Manual (ORM).

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:

Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes delete certain TS requirements which do not

meet the criteria of 10 CFR 50.36(c)(2)(ii), but identical

requirements have been added to a document (the ORM) controlled

under 10 CFR 50.59.

10 CFR 50.59 specifically prohibits changes to the facility as

described in the safety analysis report, and to procedures described

in the safety analysis report ``if the probability of occurrence or

the consequences of an accident or malfunction of equipment

important to safety previously evaluated in the safety analysis

report may be increased''. Since the conditions which limit changes

performed under 50.59 are more restrictive than the conditions which

define changes considered to involve a significant hazards

consideration, moving of a requirement from the TS to a document

which is controlled under 50.59 cannot involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Do the proposed changes create the possibility of a new or

different kind of accident from any previously evaluated?

The proposed changes delete certain TS requirements which do not

meet the criteria of 10 CFR 50.36(c)(2)(ii), but identical

requirements have been added to a document (the ORM) controlled

under 10 CFR 50.59.

10 CFR 50.59 specifically prohibits changes to the facility as

described in the safety analysis report, and to procedures described

in the safety analysis report ``if a possibility for an accident or

malfunction of a different type than any evaluated previously in the

safety analysis report may be created''. Since the conditions which

limit changes performed under 50.59 are more restrictive than the

conditions which define changes considered to involve a significant

hazards consideration, relocation of a requirement from the TS to a

document which is controlled under 50.59 cannot create the

possibility of a new or different kind of accident from any

previously evaluated.

Do the proposed changes involve a significant reduction in a

margin of safety?

The proposed changes delete certain TS requirements which do not

meet the criteria of 10 CFR 50.36(c)(2)(ii), but identical

requirements have been added to a document (the ORM) controlled

under 10 CFR 50.59.

10 CFR 50.59 specifically prohibits changes to the facility as

described in the safety analysis report, and to procedures described

in the safety analysis report if the margin of safety is reduced.

Since the conditions which limit changes performed under 50.59 are

more restrictive than the conditions which define changes considered

to involve a significant hazards consideration, relocation of a

requirement from the TS to a document which is controlled under

50.59 cannot 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.

[[Page 69338]]

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423-3698.

Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Project Director: Cynthia A. Carpenter.

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

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

Date of amendment request: July 22 and October 22, 1998.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) to reflect the licensee's

planned use of fuel supplied by Westinghouse. The Westinghouse fuel has

different design characteristics from the fuel currently in use.

Accordingly, the following changes would need to be made to the TS:

Figure 2.1.1-1, ``Reactor Core Safety Limits--Four Loops in

Operation''; various core operating parameters specified by

Surveillance Requirements 3.2.1.2, 3.2.1.3, and 3.2.2.2; Section 4.2.1,

``Fuel Assemblies''; and Section 5.6.5, ``Core Operating Limits Report

(COLR).''

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:

First Standard

Implementation of this LAR [license amendment request] would not

involve a significant increase in the probability or consequences of

an accident previously evaluated. The revised Reactor Core Safety

Limits Figure further restricts acceptable operation. Moving an

uncertainty factor from the Improved Technical Specifications to the

Core Operating Limits Report (COLR) does not exempt this factor from

regulatory restrictions. COLR parameters are generated by NRC

approved methods with the intent of ensuring that previously

evaluated accidents remain bounding. The COLR is submitted to the

NRC upon implementation of each fuel cycle or when the document is

otherwise revised. No accident probabilities or consequences will be

impacted by this LAR.

Second Standard

Implementation of this LAR would not create the possibility of a

new or different kind of accident from any previously evaluated. The

revised Reactor Core Safety Limits Figure further restricts

acceptable operation. Moving an uncertainty factor from the Improved

Technical Specifications to the COLR does not exempt this factor

from regulatory restrictions. Since the parameter in question is not

being deleted, the possibility of a new or different kind of

accident from any previously evaluated does not exist.

Third Standard

Implementation of this LAR would not involve a significant

reduction in a margin of safety. Margin of safety is related to the

confidence in the ability of the fission product barriers to perform

their design functions during and following an accident situation.

These barriers include the fuel cladding, the reactor coolant

system, and the containment system. Use of the ZIRLOTM

cladding material has been reviewed and approved in Reference 1 (as

listed in Chapter 2.1 of Topical Report DPC-NE-2009/DPC-NE-2009P,

Duke Power Company Westinghouse Fuel Transition Report).

ZIRLOTM cladding has been extensively used in

Westinghouse nuclear reactors. The changes proposed in this LAR are

necessary to ensure that the performance of the fission product

barriers (cladding) will not be impacted following the replacement

of one fuel design for another. No safety margin will be

significantly impacted.

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: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

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

422 South Church Street, Charlotte, North Carolina.

NRC Project Director: Herbert N. Berkow.

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

50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana.

Date of amendment request: November 20, 1998.

Description of amendment request: The licensee has proposed an

amendment to Facility Operating License No. NPF-47, Appendix A--

Technical Specifications (TS) Section 3.1.6, ``Control Rod Pattern.''

The proposed change will be implemented through the establishment of a

new specification added to Section 3.10, ``Special Operations.'' The

proposed specification will be TS Section 3.10.9, ``Control Rod

Pattern--Cycle 8.'' The new TS 3.10.9 is required due to a current

plant-specific configuration where 5 control rods have been inserted

into the reactor core for neutron flux suppression surrounding 2 fuel

assemblies which have been identified as having possible fuel cladding

defects. The new requirement is intended to be effective for the

remainder of the current fuel cycle (Cycle 8), and is in force when rod

withdrawal operations begin from a condition of 100% rod density to 20%

rated thermal power (RTP).

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 request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Accidents analyzed in the SAR have been examined for any impact

caused by this exception to the [Banked Position Withdrawal

Sequence] BPWS operation. The limiting event is the [Control Rod

Drive Accident] CRDA as described in SAR Sections 4.3.2 and 15.4.9.

The limit on energy addition to the fuel is 280 cal/gm as identified

in the SRP section 15.4.9. Bank Position Withdrawal Sequence is

established to reduce maximum incremental control rod worths and

thus minimize consequences resulting from an accident. The reactor

will be operated as before using BPWS. Having the current rod

configuration with 5 rods to minimize impact on the two fuel

cladding imperfections, in lieu of eight rods inoperable separated

by two cells, will not affect initiators of a Control Rod Drop

Accident. In addition, this existing rod configuration has been

analyzed and the resulting consequences continue to be bounded by

the licensing evaluations. The insertion of the identified control

rods will not affect the assumed reactivity insertion time of any

event. The location of the control rods has been reviewed by GE

using the NRC approved methodology. Operation within these limits

will ensure that the consequences of a transient or accident remain

within the acceptable limits of the evaluation. Specifically, rod

worths for the proposed configuration are bounded by the rod worths

allowed for these configurations per TS; thus, the proposed

configuration is more conservative than that allowed per TS. The

results confirm all assumed limits are maintained. The proposed

change ensures that the consequences of abnormal operation and

accidents are acceptable.

The additional Technical Specification will control the

configuration of the plant to that supported by the evaluation. If

this evaluated configuration is not supported, the plant will be

required to be placed in a configuration where the Control Rod Drop

Accident is not applicable, as the current specification requires.

The plant is therefore maintained within limits as currently

allowed. With these limits the consequences of an event are not

increased.

The probability of an accident is not affected by the proposed

Technical Specification changes since the operation of systems or

equipment that could initiate an accident are not affected.

Therefore, the proposed changes do not significantly increase the

probability or consequences of any previously evaluated accident.

(2) The request does not create the possibility of occurrence of

a new or different

[[Page 69339]]

kind of accident from any accident previously evaluated.

The proposed changes do not involve any alteration of plant

hardware or significant change in plant operation. Assuming the 5

suppression rods are bypassed in lieu of eight rods separated by two

cells does not affect event initiators or event consequences. No

plant modifications are required which would affect plant operation.

Operation with the control rod pattern in the proposed configuration

will ensure the results of a CRDA will remain within the assumptions

of the current safety analysis. The system will continue to ensure

that the limits of control rod worth remain within the assumptions

of the CRDA. The revised Technical Specifications will continue to

assure that plant operation is consistent with the assumptions,

initial conditions, and assumed power distribution and, therefore,

will not create a new type of accident.

The proposed Technical Specifications will maintain the plant in

a configuration supported by evaluation. The response to a CRDA will

be within current accepted limits and therefore no event of a

different kind has been created. The proposed Technical

Specification changes do not introduce any new modes of plant

operation nor involve new system interactions. Therefore, operation

with the 5 suppression rods inserted does not create the possibility

of an occurrence of a new or different kind of accident from any

accident previously evaluated.

(3) The request does not involve a significant reduction in a

margin of safety.

The proposed Technical Specification and the rod pattern control

system will continue to ensure the limits of control rod worth

remain within the assumptions which support the CRDA analysis of 280

cal/gm maximum energy heat addition to the fuel. This imposed limit

of 280 cal/gm provides a margin of safety from the experimental

value of approximately 330 cal/gm at which the fully molten state

for UO2 occurs. The existing rod configuration with 5

suppression rods inserted to minimize impact on the two fuel

cladding imperfections has been analyzed using NRC approved

methodology. Cycle specific evaluation has confirmed that the

consequences resulting from a CRDA continues to be bounded by the

licensing analysis for this event. Since there are no changes in the

acceptance criteria, the proposed changes will not create a

reduction in the margin of safety. These limits establish the

necessary restrictions on power operation and thereby ensure that

the core is operated within the assumptions and initial conditions

of the transient and accident analyses.

As demonstrated in the evaluation, operation within these limits

will ensure that the margin of safety will be maintained to the same

level described in the Technical Specifications Bases and the USAR

and the consequences of the postulated transient or accidents are

not increased. This limit of 280 cal/gm is not exceeded during any

transient or postulated accident. Therefore, the proposed Technical

Specifications to allow startup and continued operation in the low

power region with these control rods inserted do not involve a

significant reduction in margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

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

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

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

NRC Project Director: John N. Hannon.

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

Electric Station, Unit 3, St. Charles Parish, Louisiana.

Date of amendment request: July 2, 1998.

Description of amendment request: The proposed change will modify

the ACTION Requirements for Technical Specification (TS) 3/4.3.2 for

the Emergency Feedwater Actuation Signal (EFAS). A change to the TS

Bases Section 3/4.3.2 has been included to support this change. The

objective of this change is to add a restriction on the period of time

a channel of EFAS instrumentation can remain in the tripped condition.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No

The proposed revision to the TS changes the allowed outage time

that a channel of EFAS SGDPI [Steam Generator Differential Pressure

Instrumentation] can be in the tripped condition from a maximum of

approximately 18 months when one channel is inoperable and 92 days

when two channels are inoperable to 48 hours. If a channel were in

the tripped condition and a single failure occurred (failure of one

other channel of EFAS SGDPI), an inadvertent EFAS signal would be

generated. During a Design Basis MSLB [Main Steam Line Break] or FLB

[Feedwater Line Break] Accident, this single failure would send EFW

[Emergency Feedwater] to the faulted steam generator. The Waterford

3 safety analysis assumes that the excess Reactor Coolant System

(RCS) cooldown and return to power associated with the MSLB will be

terminated when the faulted steam generator empties. If additional

EFW were added, the RCS cooldown would be extended and the return to

power may increase.

Reducing the time that a channel of EFAS SGDPI can be placed in

the tripped condition will reduce the probability of this scenario

occurring during a Design Basis Accident. Since the allowed outage

time for a channel of EFAS SGDPI is being limited to 48 hours, this

is considered an off-normal operation and a single failure is not

required to be postulated during a Design Basis Accident in the

accident analysis. Reducing the time the channel can be placed in

the tripped condition and thus, the exposure time to this scenario,

would not be an accident initiator. The proposed change of being

more conservative relative to allow[ed] outage time in the tripped

condition will not affect the assumptions, design parameters, or

results of any accident previously evaluated.

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.

The proposed change does not alter the design or configuration

of the plant. The proposed change provides a more conservative

allowed outage time for the channel to be in the tripped condition.

There has been no physical change to plant systems, structures or

components nor will the proposed change reduce the ability of any of

the safety-related equipment required to mitigate Anticipated

Operational Occurrences or accidents. The configuration required by

the proposed specification is permitted by the existing

specification.

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 provides a more conservative allowed outage

time for the channel to be in the tripped condition. By reducing the

allowed outage time, the probability is reduced that a single

failure (failure of one channel of EFAS SGDPI with one channel in

the tripped condition) would occur that would send EFW to the

faulted steam generator. Therefore, the only change to the margin of

safety would be an increase. Since the allowed outage time for a

channel of EFAS SGDPI is being limited to 48 hours, this is

considered an off-normal operation and a single failure is not

required to be postulated during a Design Basis Accident in the

accident analysis. The proposed changes do not affect the limiting

conditions for operation or their bases.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

[[Page 69340]]

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 NW., Washington, DC 20005-3502.

NRC Project Director: John N. Hannon.

Florida Power and Light Company, et al., Docket No. 50-389, St.

Lucie Plant, Unit No. 2, St. Lucie County, Florida.

Date of amendment request: December 31, 1997, as supplemented

November 25, 1998.

Description of amendment request: The proposed amendment will

revise the St. Lucie Unit 2 Technical Specifications to permit an

increase in the allowed Spent Fuel Pool (SFP) storage capacity.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Analyses to support the proposed fuel pool capacity increase

have been developed using conservative methodology. The analysis of

the potential accidents summarized below has shown that there is no

significant increase in the consequences of any accident previously

analyzed. A review of relevant plant operations has also

demonstrated that there is no significant increase in the

probability of occurrence of any accident previously analyzed. This

conclusion is also discussed below.

Previously evaluated accidents that were examined for this

proposed license amendment include: Fuel Handling Accident, Spent

Fuel Cask Drop Accident, and Loss of all Fuel Pool Cooling.

There will be no change in the mode of plant operation or in the

availability of plant systems as a result of this proposed change;

the systems interfacing with the spent fuel pool have previously

encountered borated pool water and are designed to interact with

irradiated spent fuel and remove the residual heat load generated by

isotopic decay. The proposed amendment does not require a change in

the maintenance interval or maintenance scope for the fuel pool

cooling system or for the spent fuel cask crane. The frequency of

cask handling operations and the maximum weight carried by the crane

is not increased as a result of the proposed license amendment.

Thus, there will be no increase in the probability of a loss of fuel

pool cooling or in the probability of a failure of the cask crane as

a result of the proposed amendment.

There will not be a significant increase in the frequency of

handling discharged assemblies in the fuel pool as a result of this

change; any handling of fuel in the spent fuel pool will continue to

be performed in borated water. If the license amendment is approved,

there will be a one-time repositioning of certain discharged

assemblies stored in the fuel pool to comply with the revised

positioning requirements, but the increased pool storage capacity

will permit the deferral of spent fuel handling associated with cask

loading operations. Fuel manipulation during the repositioning

activity will be performed in the same manner as for fuel placed in

the spent fuel pool during refueling outages. There will be no

changes in the manner of handling fuel discharged from the core as a

result of refueling; administrative controls will continue to be

used to specify fuel assembly placement requirements. The relative

positions of Region I and Region II storage locations will remain

the same within the fuel pool. Therefore, the probability of a fuel

handling accident has not been significantly increased.

The consequences of a fuel handling accident have been

evaluated. The radioactive release consequences of a dropped fuel

assembly are not affected by the proposed increase in fuel pool

storage capacity. They remain bounded by the results of calculations

performed to justify the existing St. Lucie Unit 2 fuel storage

racks and burnup limits. At the limiting fuel assembly burnup,

radioactive releases from a dropped assembly would be only a small

fraction of NRC guidelines. The input parameters employed in

analyzing this event are consistent with the current values of fuel

enrichment, discharge burnup and uranium content used at St. Lucie

Unit 2 and with future use of the ``value-added'' fuel pellet

design. Thus, the consequences of the fuel assembly drop accident

would not be significantly increased from those previously

evaluated.

The capability of the fuel pool cooling system to handle the

increased number of discharged assemblies has been examined. The

impact of a total loss of spent fuel pool cooling flow on available

equipment recovery time and on fuel cladding integrity has also been

evaluated. For the limiting full core discharge, sufficient time

remains available to restore cooling flow or to provide an alternate

makeup source before boiloff results in a fuel pool water level less

than that needed to maintain acceptable radiation dose levels.

Analysis has shown that in the event of a total loss of fuel pool

cooling fuel cladding integrity is maintained. Therefore, the

consequences of a loss of fuel pool cooling event, including the

effect of the proposed increase in fuel pool storage capacity, have

not been significantly increased from previously analyzed results

for this type of accident.

The analysis of record pertaining to the radiological

consequences of the hypothetical drop of a loaded spent fuel cask

just outside the Fuel Handling Building was examined to determine

the impact of the increased fuel storage capacity on this accident's

results. The results of the previously performed analysis were

determined to bound the conditions described by the proposed license

amendment, thus the consequences of the cask drop accident would not

be significantly increased as a result of this change.

It is concluded that the proposed amendment to increase the

storage capacity of the St. Lucie Unit 2 spent fuel pool will not

involve a significant increase in the probability or consequences of

any accident previously evaluated.

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

new or different type of accident from any accident previously

evaluated.

In this license amendment FPL [Florida Power & Light Co.]

proposes to credit the negative reactivity associated with a portion

of the soluble boron present in the spent fuel pool. Soluble boron

has always been present in the St. Lucie Unit 2 spent fuel pool; as

such the possibility of an inadvertent fuel pool dilution has always

existed. However, the spent fuel pool dilution analysis demonstrates

that a dilution of the Unit 2 spent fuel pool which could increase

the pool keff to greater than 0.95 is not a credible

event. Neither implementation of credit for the reactivity of fuel

pool soluble boron nor the proposed increase in the fuel pool

storage capacity will create the possibility of a new or different

type of accident at St. Lucie Unit 2.

An examination of the limiting fuel assembly misload has

determined that this would not represent a new or different type of

accident. None of the other accidents examined as a part of this

license submittal represent a new or different type of accident;

each of these situations has been previously analyzed and determined

to produce acceptable results.

The proposed license amendment will not result in any other

changes in the mode of spent fuel pool operation at St. Lucie Unit 2

or in the method of handling irradiated nuclear fuel. The spatial

relationship between the fuel storage racks and the cask crane range

of motion is not affected by the proposed change.

As a result of the evaluation and supporting analyses, FPL has

determined that the proposed fuel pool capacity increase does not

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

any accident previously evaluated.

3. The proposed amendment will not involve a significant

reduction in the margin of safety.

FPL has determined, based on the nature of the proposed license

amendment that the issue of margin of safety, when applied to this

fuel pool capacity increase, should address the following areas:

1. Fuel Pool reactivity considerations

2. Fuel Pool boron dilution considerations

3. Thermal-Hydraulic considerations

4. Structural loading and seismic considerations

The Technical Specification changes proposed by this license

amendment, the proposed spent fuel pool storage

[[Page 69341]]

configuration and the existing Technical Specification limits on

fuel pool soluble boron concentration provide sufficient safety

margin to ensure that the array of fuel assemblies stored in the

spent fuel pool will always remain subcritical. The revised spent

fuel storage configuration is based on a Unit 2 specific criticality

analysis performed using methodology consistent with that approved

by the NRC. Additionally, the soluble boron concentration required

by current Technical Specifications ensures that the fuel pool

keff will be always be maintained substantially less than

0.95.

The Unit 2 criticality analysis established that the

keff of the spent fuel pool storage racks will be less

than 1.0 with no soluble boron in the fuel pool water, including the

effect of all uncertainties and tolerances. Credit for the soluble

boron actually present is used to offset uncertainties, tolerances,

off-normal conditions and to provide margin such that the spent fuel

pool keff is maintained less than or equal to 0.95. FPL

has also demonstrated that a decrease in the fuel pool boron

concentration such that keff exceeds 0.95 is not a

credible event.

Current Technical Specifications require that the fuel pool

boron concentration be maintained greater than or equal to 1720 ppm.

This boron value is substantially in excess of the 520 ppm required

by the uncertainty and reactivity equivalencing analyses discussed

in this evaluation and the 1266 ppm value required to maintain

keff less than or equal to 0.95 in the presence of the

most adverse mispositioned fuel assembly.

The St. Lucie Unit 2 fuel pool boron concentration will continue

to be maintained significantly in excess of 1266 ppm; the proposed

license amendment will not result in changes in the mode of

operation of the refueling water tank (RWT) or in its use for makeup

to the fuel pool. Thus, operation of the spent fuel pool following

the proposed change, combined with the existing fuel pool boron

concentration Technical Specification limit of 1720 ppm, will

continue to ensure that keff of the fuel pool will be

substantially less than 0.95.

Even if this not-credible dilution event was to occur, no

radiation would be released; the only consequence would be a

reduction of shutdown margin in the fuel pool. The volume of

unborated water required to dilute the fuel pool to a

keff of 0.95 is so large (in excess of 358,900 gallons to

dilute the fuel pool to 520 ppm boron) that only a limited number of

water sources could be considered potential dilution sources. The

likelihood that this level of water use could remain undetected by

plant personnel is extremely remote.

In meeting the acceptance criteria for fuel pool reactivity, the

proposed amendment to increase the storage capacity of the existing

fuel pool racks does not involve a significant reduction in the

margin of safety for nuclear criticality.

Calculations of the spent fuel pool heat load with an increased

fuel pool inventory were performed using ANSI/ANS-5.1-1979

methodology. This method was demonstrated to produce conservative

results through benchmarking to actual St. Lucie Unit 2 fuel pool

conditions and by comparison of its results to those generated by a

calculation using Auxiliary Systems Branch Technical Position 9-2

methodology. Conservative methods were also used to demonstrate fuel

cladding integrity is maintained in the absence of cooling system

forced flow. The results of these calculations demonstrate that, for

the limiting case, the existing fuel pool cooling system can

maintain fuel pool conditions within acceptable limits with the

increased inventory of discharged assemblies.

Therefore, the proposed change does not result in a significant

reduction in the margin of safety with respect to thermal-hydraulic

or spent fuel cooling considerations.

The primary safety function of the spent fuel pool and the fuel

storage racks is to maintain discharged fuel assemblies in a safe

configuration for all environments and abnormal loadings, such as an

earthquake, a loss of pool cooling or a drop of a spent fuel

assembly during routine spent fuel handling. The proposed increase

in spent fuel inventory on the fuel pool and the existing storage

racks have been evaluated and show that relevant criteria for fuel

rack stresses and floor loadings have been met and that there has

been no significant reduction in the margin of safety for these

criteria.

The NRC staff has reviewed the licensee's analysis and the changes

proposed in the November 25, 1998 supplement to the original submittal

and based on this review, it appears that the three standards of

50.92(c) continue to be satisfied. Therefore, the NRC staff proposes to

determine that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Indian River Junior College

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

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

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

NRC Project Director: Frederick J. Hebdon.

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

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

Date of amendment request: October 27, 1998.

Description of amendment request: The licensee proposed to change

Technical Specification (TS) 6.3, Facility Staff Qualifications, in

order to incorporate qualifications for the Multi-Discipline

Supervisor. The current TS requires that plant staff meet the

requirements of the American National Standards Institute (ANSI) N18.1-

1971, which requires non-licensed supervisors to have a high school

diploma or equivalent and a minimum of 4 years experience in the craft

or discipline they supervise. The proposed change requires the Multi-

Discipline Supervisor to have, (1) a high school diploma or equivalent,

(2) a minimum of 4 years of related technical experience, which shall

include 3 years of power plant experience of which one year is at a

nuclear power plant, and (3) completed the Multi-Discipline Supervisor

training program.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed changes are administrative in nature addressing

personnel qualification issues. The Multi-Discipline Supervisor

(MDS) position will be filled with personnel who are experienced in

one or more technical disciplines (maintenance, operations,

engineering, or other related technical discipline). Fundamental

working knowledge of tasks being performed will be acquired through

the MDS initial training program. The training concentrates on

developing the skills and knowledge of an MDS to safely oversee

tasks for multi-discipline work teams. Therefore, four years

experience in any related technical discipline or disciplines

combined with the MDS training program provide adequate technical

knowledge for proper job oversight. These proposed changes will not

involve a significant increase in the probability or consequences of

an accident previously evaluated because they do not affect

assumptions contained in plant safety analyses, the physical design

and/or operation of the plant, nor do they affect Technical

Specifications that preserve safety analysis assumptions. Therefore,

the proposed changes do not affect the probability or consequences

of accidents previously analyzed.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The changes being proposed are administrative in nature and do

not affect assumptions contained in plant safety analyses, the

physical design and/or modes of plant operation defined in the

facility operating license, or Technical Specifications that

preserve safety analysis assumptions. These changes address

qualification requirements for the MDS position. Since the proposed

changes do not change the qualifications for those individuals

responsible for the actual licensed operation of the facility,

operation of the facility in accordance with the proposed amendments

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

accident from any accident

[[Page 69342]]

previously evaluated. No new failure mode is introduced due to the

administrative changes since the proposed changes do not involve the

addition or modification of equipment nor do they alter the design

or operation of affected plant systems, structures, or components.

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

amendments would not involve a significant reduction in a margin of

safety.

The operating limits and functional capabilities of the affected

systems, structures, and components are unchanged by the proposed

amendments. The proposed changes to add the MDS position have

management and administrative controls associated with the required

qualification requirements. The Turkey Point Technical

Specifications will ensure that any individual filling the MDS

position has the requisite education, experience, and training. As a

result, operation of the facility in accordance with the proposed

changes would not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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

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

NRC Project Director: Frederick J. Hebdon.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear

Generating Station, Ocean County, New Jersey.

Date of amendment request: November 5, 1998.

Description of amendment request: The proposed Technical

Specification change will modify the safety limits and surveillances of

the LPRM and APRM systems and related Bases pages to ensure the APRM

channels respond within the necessary range and accuracy and to verify

channel operability. In addition, an unrelated change to the Bases of

Specification 2.3 is included to clarify some ambiguous language.

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 technical specification changes to the limits

and surveillance requirements of the LPRM and APRM systems are

provided to ensure the APRM channels respond within the necessary

range and accuracy and to verify channel operability. If one or more

monitored parameters exceeded their specified limits, the RPS

initiates a reactor scram signal to preserve the integrity of the

fuel cladding and the Reactor Coolant System and minimize the energy

that must be absorbed following a loss of coolant accident.

Therefore, the probability of occurrence or the consequences of an

accident previously evaluated in the [safety analysis report] SAR

will not increase as a result of these changes.

2. The proposed technical specification changes to the limits

and surveillance requirements of the LPRM and APRM systems are

provided to ensure the APRM channels respond within the necessary

range and accuracy and to verify channel operability. The proposed

changes are designed to ensure the APRM system responds in a manner

that ensures the safety limits, limiting safety system settings,

limiting conditions for operations, as well as design parameters for

the APRM system and individual components are continuously met.

Therefore, the proposed activity does not create the possibility for

an accident or malfunction of a different type than any previously

identified in the SAR.

3. The proposed change does not involve a significant reduction

in the margin of safety. When the APRMs exceed their specified

limits, the RPS initiates a reactor scram signal to preserve the

integrity of the fuel cladding and the Reactor Coolant System and

minimize the energy that must be absorbed following a loss of

coolant accident. The proposed changes are designed to assure the

APRM system responds in a manner that ensures the safety limits,

limiting safety system settings, limiting conditions for operations,

as well as design parameters for the APRM system and individual

components are continuously met. Therefore, the margin of safety

will not be reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

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

NRC Project Director: Cecil O. Thomas.

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

Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania.

Date of amendment request: November 25, 1998.

Description of amendment request: The proposed amendment will

change the surveillance specification for Once Through Steam Generator

(OTSG) inservice inspections for TMI-1 Cycle 13 refueling outage

examinations which would be applicable for the next operating cycle

only, Operating 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:

A. The proposed changes do not represent a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed flaw disposition strategy, based on measurable eddy

current parameters of axial and circumferential extent for Inside

Diameter (ID) Initiated Inter-Granular Attack (IGA), will continue

to provide high confidence that unacceptable flaws that do not have

the required structural integrity to withstand a postulated MSLB

[main steam line break] are removed from service. The axial and

circumferential length limits for eddy current ID degradation

indications meet the Draft Regulatory Guide 1. 121 * * * acceptance

criteria for margin to failure for MSLB-applied differential

pressure and axial tube loads. The capability for detection of flaws

is unaffected; and the identification of tubes that should be

repaired or removed from service is maintained. The operation of the

OTSGs or related structures, systems, or components is otherwise

unaffected. Therefore, neither the probability nor consequences of

[an] SGTR [steam generator tube rupture] is significantly increased

either during normal operation or due to the limiting loads of [an]

MSLB accident.

Neither the change in voltage normalization for the eddy current

examinations, nor the administrative change in clarification of the

reporting requirements, as described above, could significantly

affect the probability of occurrence or consequences of any accident

previously evaluated. These changes are administrative only.

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

or different kind of accident from any accident previously evaluated

because there are no hardware changes involved nor changes to any

operating practices. These changes involve only the OTSG tube

inservice inspection surveillance requirements, which could only

affect the potential for OTSG primary-to-secondary leakage. The

proposed changes continue to impose flaw length limits for ID IGA to

assure tube structural and leakage integrity, as confirmed by 12R

(and post 12R) tube pull sample examinations and pressure testing.

In addition, neither the change in voltage normalization for the

eddy current examinations nor the administrative change in the

description of the reporting requirements, as described above, could

possibly create the possibility of an accident

[[Page 69343]]

of a new or different type from any previously evaluated. These

changes are included only to modify the plant's eddy current

normalization to the industry standard, and clarify the reporting

period for submittal of the OTSG inspection results to the NRC

[Nuclear Regulatory Commission]. Therefore, these changes do not

create the potential for any other kind of accident different from

those that have been evaluated.

C. These proposed changes do not involve a significant reduction

in a margin of safety because the margins of safety defined in Draft

Regulatory Guide 1. 121 * * * are retained. The probability of

detecting degradation is unchanged since the bobbin coil eddy

current methods will continue to be the primary means of initial

detection and the probability of leakage from any indications left

in service remains acceptably small. The strategy for dispositioning

ID initiated IGA will continue to provide a high level of confidence

that tubes exceeding the allowable limits for tube integrity are

repaired or removed from service.

In addition, neither the change in voltage normalization for the

eddy current examinations nor the administrative change in the

description of the reporting requirements, as described above, could

significantly affect a margin of safety. These changes are

administrative in nature and are included only to align TMI-1's

voltage normalization to the industry standard, and clarify the

reporting period, respectively.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

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

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

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

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

NRC Project Director: Cecil O. Thomas.

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

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

Connecticut.

Date of amendment request: November 10, 1998.

Description of amendment request: The proposed changes would modify

Technical Specifications 3.3.1.1, ``Reactor Protective

Instrumentation,'' and 3.3.2.1, ``Engineered Safety Feature Actuation

System Instrumentation'' to restrict the time a reactor protection or

engineered safety feature actuation channel can be in the bypass

position to 48 hours, from an indefinite period of time. Most of these

proposed changes were originally submitted in a letter dated May 14,

1998. The licensee withdrew its original request and submitted a new

request in its November 10, 1998, letter.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

Company] has reviewed the proposed changes and has concluded that

they do not involve a significant hazards consideration (SHC). The

basis for this conclusion is that the three criteria of

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

involve an SHC because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to restrict the time [* * *] reactor

protection or engineered safety feature actuation channels can be in

the bypass position to 48 hours, from an indefinite period of time,

has no effect on the design of the Reactor Protection System (RPS)

or the Engineered Safety Feature Actuation System (ESFAS) and does

not affect how these systems operate. In addition, this will

minimize the susceptibility of these systems to the remote

possibility of fault propagation between channels. However, this

proposed change will require an inoperable pressurizer high pressure

reactor protection channel to be placed in the tripped condition

within 48 hours. With a pressurizer pressure channel in the tripped

condition, the high failure of a second pressurizer pressure channel

would initiate a reactor trip and open both pressurizer power

operated relief valves (PORVs). Opening the pressurizer PORVs would

result in an undesired loss of primary coolant. Thus, this change

will increase the probability of occurrence of a previously

evaluated accident. However, this would not place the plant in an

unanalyzed condition since FSAR [Final Safety Analysis Report]

Section 14.6.1 analyzes the inadvertent opening of both PORVs, the

release of reactor coolant can be terminated by closure of the PORV

block valves from the control room, and the Emergency Operating

Procedures provide guidance on how to address this situation.

Therefore, this change does not significantly increase the

probability or consequences of an accident previously evaluated.

The proposed change to increase the time a second RPS or ESFAS

channel can be removed from service (from 2 hours to 48 hours),

provided one of the inoperable channels is placed in the tripped

condition, has no effect on the design of the RPS or ESFAS and does

not affect how these systems operate. These systems will still

function as designed to mitigate design basis accidents. However,

this change will also impact the probability of occurrence of a

previously evaluated accident since it will allow a second

pressurizer high pressure reactor protection channel to be placed in

the tripped condition for 48 hours instead of the current 2 hour

time limit. The impact of this change is bounded by the proposed

change to require an inoperable pressurizer high pressure reactor

protection channel to be placed in the tripped condition after 48

hours as previously discussed. Therefore, this change does not

significantly increase the probability or consequences of an

accident previously evaluated.

The proposed change to apply a more restrictive action statement

to the loss of turbine load reactor trip function has no effect on

the design of this trip function and does not affect how this trip

function operates. Also, this trip function is not assumed to

operate to mitigate any design basis accident. Therefore, this

change does not significantly increase the probability or

consequences of an accident previously evaluated.

The proposed change to require a channel calibration every 18

months for the loss of turbine load reactor trip function and for

the wide range logarithmic neutron flux monitors has no effect on

the design of either the loss of turbine load reactor trip function

or the wide range logarithmic neutron flux monitors. Also, neither

of these are assumed to operate to mitigate any design basis

accident. Therefore, this change does not significantly increase the

probability or consequences of an accident previously evaluated.

The proposed change to exclude the neutron detectors from the

channel calibration requirement has no effect on the design of the

neutron detectors and has no significant effect on how these

detectors operate. The detectors are passive devices with minimal

drift. In addition, slow changes in the sensitivity of the linear

power range flux detectors is compensated for by performing the

daily calorimetric calibration and the monthly calibration using the

incore detectors. These detectors will still function as designed to

mitigate design basis accidents. Therefore, this change does not

significantly increase the probability or consequences of an

accident previously evaluated.

The proposed change to add the license amendment numbers to

Technical Specification Page 3/4 3-9 will not result in a technical

change to the Millstone Unit No. 2 Technical Specifications. The RPS

will continue to function as before. Therefore, this change does not

significantly increase the probability or consequences of an

accident previously evaluated.

The proposed change to correct the surveillance requirement

referenced in an action statement has no effect on the design of the

ESFAS and does not affect how this system operates. The ESFAS will

still function as designed to mitigate design basis accidents.

Therefore, this change does not significantly increase the

probability or consequences of an accident previously evaluated.

The proposed change to add a reference to the reactor coolant

pump low speed reactor trip function to a note that states this trip

[[Page 69344]]

may be bypassed Q, Nuclear Enthalpy Rise Hot Channel Factor,

LOCA PCT [Peak Clad Temperature], peak local power density, or any

other margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Missouri-

Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

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

& Trowbridge, 2300 N Street, NW, Washington, DC 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: November 18, 1998.

Description of amendment request: The proposed amendments would

make changes to the North Anna Power Station (NAPS), Unit 1 and 2,

Technical Specifications (TS) Surveillance Requirement (SR) 4.7.13.1,

``Groundwater Surveillance Requirements'' and related Table 3.7-6,

``Allowable Groundwater Levels--Service Water Reservoir.'' The change

in the SR requests that the measuring device numbers assigned to

piezometers be eliminated from the TS SR in order to avoid redundancy,

and eliminate confusion as well as the need to initiate TS changes

whenever new piezometers are added, older devices are replaced or

abandoned in-place. The proposed change in groundwater threshold levels

will raise the allowable groundwater levels to those consistent with

the allowable levels in the ``Stability of Service Water Reservoir

(SWR) Slope Under Increased Phreatic Surface'' calculations.

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, which is

presented below:

Specifically, operation of the North Anna Power Station in

accordance with the proposed TS Change Request will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, since: (a)

removing non-safety related SWR piezometer device numbers from the

TS and raising TS allowable groundwater surface threshold elevation

levels in the southeast section of the SWR will have no effect on

the way the safety-related Service Water System was designed to

operate, (b) Periodic Test Procedures will continue to identify all

open-tube piezometers and require that they be monitored in order to

obtain as much information as possible regarding changing

groundwater levels, (c) sufficient redundancy will continue to exist

since at least two (2) open-tube (standpipe-type) piezometers, not

subject to mechanical failure, have been installed in each of the

three (3) SWR zones to meet the TS Surveillance Requirement that

``at least one measurement per zone be available'' and (d) recent

calculations have confirmed that raising the allowable water level

in the southeast section of the SWR will not affect the stability of

the SWR dike as indicated in the original design basis calculation.

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

from any accident previously evaluated, since: (a) the frequency of

piezometer monitoring and the intent of monitoring groundwater

surface threshold elevations in order to maintain stability of the

SWR slope have not changed, (b) no physical modification to the

plant or new mode of plant operation is involved, (c) changes are

consistent with the assumptions made in the Safety Analyses and

original design basis calculation and (d) failure of the SWR dike

and ensuing loss of service water was the most serious accident

postulated and considered credible. Operation of the SWR is not

being changed. Therefore, a new or different kind of accident is

[not] created by the change in groundwater level. In addition, since

both the SWR and Lake Anna reservoir provide redundant sources of

service water, failure of the SWR is not considered as a credible

accident.

3. Involve a significant reduction in a margin safety, since:

(a) increasing the allowable phreatic surface in the SE section of

the SWR dike will not lower the factor of safety with respect to the

stability of the SWR as defined by the original design basis

calculation, (b) the margin to failure of the SWR dike has been

proven by calculation to have not been reduced as defined by the

original design basis calculation and (c) subject changes will not

impact the performance of structures, systems or components relied

upon for accident mitigation or any safety analysis assumptions,

therefore the margin of safety is not changed by the proposed

[change] in groundwater level at the SWR.

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: Donald P. Irwin, Esq., Hunton and Williams,

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

23219.

NRC Project Director: Herbert N. Berkow.

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

[[Page 69350]]

Commission's rules and regulations. The Commission has made appropriate

findings as required by the Act and the Commission's rules and

regulations in 10 CFR Chapter I, which are set forth in the license

amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

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

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

the particular facilities involved.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

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

Carolina.

Date of application for amendment: February 27, 1997, as

supplemented August 24, 1998.

Brief description of amendment: This amendment changes Technical

Specification (TS) 3/4.4.5, ``Steam Generators,'' by adding sleeve

installation as an alternative to tube plugging for repairing degraded

steam generators.

Date of issuance: November 23, 1998.

Effective date: November 23, 1998.

Amendment No.: 85.

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

Technical Specifications.

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

17225).

The August 24, 1998, supplemental letter provided clarifying

information only, and did not change the initial no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

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

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

Power Station, Unit 1, Rock Island County, Illinois.

Date of application for amendment: August 14, 1998, as supplemented

by letters dated October 13 and November 23, 1998.

Brief description of amendment: The amendment changes the Quad

Cities Technical Specifications (TS) to reflect the use of Siemens

Power Corporation ATRIUM-9B fuel. Specifically the amendment

incorporates 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.

Date of issuance: December 3, 1998.

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

Amendment No.: 182.

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

TSs. Public comments requested as to proposed no significant hazards

consideration: Yes (63 FR 59588 dated November 4, 1998). This notice

provided an opportunity to submit comments on the Commission's proposed

no significant hazards consideration determination. No comments have

been received. The notice also provided for an opportunity to request a

hearing by December 4, 1998, but indicated that if the Commission makes

a final no significant hazards consideration determination any such

hearing would take place after issuance of the amendment.

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

exigent circumstances, and final no significant hazards consideration

determination are contained in a Safety Evaluation dated December 3,

1998.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

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

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

Date of application for amendment: October 31, 1997, as

supplemented December 13, 1997, February 27 and April 24, 1998.

Brief description of amendment: The amendment proposed to revise

the Final Safety Analysis Report (FSAR) to reflect changes to the

credited methodology for boron precipitation prevention, as approved by

the NRC.

Date of issuance: November 30, 1998.

Effective date: November 30, 1998.

Amendment No.: 171.

Facility Operating License No. DPR-72: Amendment revised the

Operating License to reflect the change to the FSAR.

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

FR 60731). The supplemental letters contained clarifying information

that did not change the original no significant hazards consideration

determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 30, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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: October 29, 1998.

Brief description of amendment: The amendment revised the

terminology used in the St. Lucie Plant Technical Specifications (TS)

relative to the implementation and automatic removal of certain

protection system trip bypasses to ensure that the meaning of explicit

terms used in the TS are consistent with the intent of the stated

requirements.

Date of Issuance: November 24, 1998.

Effective Date: November 24, 1998.

Amendment No.: 98.

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

Date of initial notice in Federal Register: November 5, 1998 (63 FR

59809).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 24, 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.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear

[[Page 69351]]

Generating Station, Ocean County, New Jersey.

Date of application for amendment: July 21, 1998.

Brief description of amendment: The amendment (1) revises Technical

Specification (TS) 6.2.2.2(a) to provide flexibility to accommodate

unexpected absence of on-duty shift crew members, (2) eliminates

reference to the Manager, Plant Operations in Specification 6.2.2.2(j)

as the position has been eliminated, (3) reduces the maximum time in

which to forward audit reports to the responsible manager from 60 days

to 30 days, (4) replaces the term ``Vice President'' with the term

``Corporate Officer'' in several places in Section 6, and (5) corrects

several typographical errors.

Date of Issuance: November 30, 1998.

Effective date: November 30, 1998, to be implemented within 30 days

Amendment No: 203.

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

Technical Specifications.

Date of initial notice in Federal Register: August 26, 1998 (63 FR

45525).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated November 30, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan.

Date of application for amendments: October 8, 1998.

Brief description of amendments: The amendments would revise the

Technical Specification Section 3.4.1.3, ``Reactor Coolant System--

Shutdown,'' and its associated bases to provide separate requirements

for the Reactor Coolant system in MODE 4, MODE 5 with the reactor

coolant loops filled, and MODE 5 with the reactor coolant loops not

filled.

Date of issuance: November 27, 1998.

Effective date: November 27, 1998, with full implementation within

30 days.

Amendment Nos.: 224 and 208.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1998 (63 FR

57322).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 27, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, MI 49085.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile

Point Nuclear Station Unit No. 1, Oswego County, New York.

Date of application for amendment: June 19, 1998, as supplemented

November 6, 1998.

Brief description of amendment: This amendment changes Technical

Specification 3.2.2 and the associated Bases to update pressure-

temperature operating curves and tables for continued plant operation

up to 28 effective full-power years.

Date of issuance: November 25, 1998.

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

core operation exceeds 18 effective full-power years.

Amendment No.: 164.

Facility Operating License No. DPR-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: July 29, 1998 (63 FR

40557)

The November 6, 1998, supplemental letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

November 25, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York.

Date of application for amendment: November 25, 1998, as

supplemented November 27, 1998.

Brief description of amendment: This change adds a note to certain

specific containment isolation valves listed in Table 4.4-1. The note

permits the licensee to operate Indian Point Unit 3 for the remainder

of the current cycle (Cycle 10) without pneumatic leakage rate testing

of these isolation valves. These valves have been leakage rate tested

in the past using water pressurized with nitrogen gas. Without this

emergency amendment, there would have had to delay its resumption of

plant operation at power until the Technical Specifications required

test was performed.

Date of issuance: November 27, 1998.

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

immediately.

Amendment No.: 184.

Facility Operating License No. DPR-64: Amendment revised the

Technical Specifications. The Commission's related evaluation of the

amendment, finding of emergency circumstances, and final determination

of no significant hazards consideration, are contained in a Safety

Evaluation dated November 27, 1998.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa, Director.

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: August 3, 1998, as supplemented

October 20, 1998.

Brief description of amendment: The amendment provides for

application of the existing minimum critical power ratio safety limit

to Cycle 14 operation.

Date of issuance: November 25, 1998.

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

30 days.

Amendment No.: 246.

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

Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48264).

The October 20, 1998, supplemental letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 25, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

Southern California Edison Company, et al., Docket No. 50-362, San

Onofre Nuclear Generating Station, Unit No. 3, San Diego County,

California.

Date of application for amendment: September 22, 1998.

[[Page 69352]]

Brief description of amendment: The proposed amendment would modify

the Technical Specifications (TS) to change the parameter used to

establish and remove the bypasses for high reactor power trips. The

parameter would be changed from the current ``THERMAL POWER'' to

logarithmic power.

Date of issuance: November 23, 1998.

Effective date: November 23, 1998.

Amendment Nos.: 136.

Facility Operating License No. NPF-15: The amendments revised the

Technical Specifications.

Date of initial notice in Federal Register: October 21, 1998 (63 FR

56259).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

California, P. O. Box 19557, Irvine, California 92713.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

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

Alabama.

Date of application for amendments: June 12 and August 14, 1998

(TS-390).

Brief description of amendments: Changes the technical

specifications (TS) to accommodate surveillance intervals to be

compatible with a 24-month fuel cycle.

Date of issuance: November 30, 1998.

Effective date: November 30, 1998.

Amendment Nos.: 235, 255, 215.

Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:

Amendments revised the TS.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48269).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 30, 1998.

No significant hazards consideration comments received: None.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611.

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 21, 1996 (TS 96-03).

Brief description of amendments: The amendments revise the SQN

Technical Specification (TS) 3.7.1.3 to extend the limiting condition

for operation of the condensate storage tanks to Mode 4 when steam

generator is relied upon for heat removal.

Date of issuance: November 19, 1998.

Effective date: As of the date of issuance to be implemented no

later than 45 days after issuance.

Amendment Nos.: 238 and 228.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the TSs.

Date of initial notice in Federal Register: October 9, 1996 (61 FR

52967).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 19, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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: April 30, 1998 (TS 98-01).

Brief description of amendments: The amendments revise the SQN

Technical Specification Surveillance Requirement 4.4.3.2.1.b by

changing the mode requirement to allow power-operated relief valve

stroke testing in Modes 3, 4, and 5 with a steam bubble in the

pressurizer rather than only in Mode 4.

Date of issuance: November 19, 1998.

Effective date: As of the date of issuance to be implemented no

later than 45 days after issuance.

Amendment Nos.: 239 and 229.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the technical specifications.

Date of initial notice in Federal Register: July 15, 1998 (63 FR

38204).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 19, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear

Plant, Unit 1, Rhea County, Tennessee.

Date of application for amendment: May 6, as supplemented June 5,

1998.

Brief description of amendment: The requested changes would allow

an increase in the limit, up to 5.0 percent, for the U-235 enrichment

of new (unirradiated) fuel stored in the new fuel storage racks and

limit the fuel storage locations to assure that k-effective values are

met.

Date of issuance: December 1, 1998.

Effective date: December 1, 1998.

Amendment No.: 15.

Facility Operating License No. NPF-90: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: August 12, 1998 (63 FR

43214).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 1, 1998.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear,

Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No. 50-

440 Perry Nuclear Power. Plant, Unit 1, Lake County, Ohio.

Date of application for amendment: September 3, 1998.

Brief description of amendment: This amendment revised Technical

Specification 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,''

by increasing the Division 3 Diesel Generator fuel oil level

requirements to account for (1) a rounding error in the calculation,

and (2) the unusable volume due to vortex formation at the eductor

suction nozzle located in the fuel oil storage tank.

Date of issuance: November 23, 1998.

Effective date: November 23, 1998.

Amendment No.: 94.

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

Technical Specifications.

Date of initial notice in Federal Register: October 7, 1998 (63 FR

53960).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear,

Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No. 50-

440 Perry Nuclear Power Plant, Unit 1, Lake County, Ohio.

Date of application for amendment: August 28, 1997.

Brief description of amendment: This amendment revised Pressure-

[[Page 69353]]

Temperature (P/T) Limits contained in Technical Specification 3.4.11 as

a result of the Reactor Vessel Material Surveillance Program

Requirements contained in Appendix H of 10 CFR Part 50.

Date of issuance: December 2, 1998.

Effective date: December 2, 1998.

Amendment No.: 95.

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

Technical Specifications.

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

FR 61846).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 2, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin.

Date of application for amendment: April 15, 1998 as supplemented

by letters dated August 13, 1998, September 28, 1998, and November 24,

1998.

Brief description of amendment: The amendment incorporates changes

to TS 2.1, ``Safety Limits'' and TS 3.10, ``Control Rod and Power

Distribution Limits.'' These changes revise the power distribution

peaking factor limits and limits operating parameters related to the

Minimum Departure from Nucleate Boiling Ratio (MDNBR) in support of

cycle 23 fuel and reload changes. A change associated with the fuel and

reload changes, is the removal, from the current licensing basis, of

the fuel pool turbine missile hazards analysis

Date of issuance: December 2, 1998.

Effective date: December 2, 1998.

Amendment No.: 142.

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 5, 1998 (63FR25120

).

The supplemental submittals did not affect the initial

determination of no significant hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 2, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

Dated at Rockville, Maryland, this 9th day of December 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

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

Nuclear Reactor Regulation.

[FR Doc. 98-33206 Filed 12-15-98; 8:45 am]

BILLING CODE 7590-01-P