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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from December 18, 1998, through December 31,

1998. The last biweekly notice was published on December 30, 1998 (63

FR 71962).

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 February 12, 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

[[Page 2244]]

or fact. Contentions shall be limited to matters within the scope of

the amendment under consideration. The contention must be one which, if

proven, would entitle the petitioner to relief. A petitioner who fails

to file such a supplement which satisfies these requirements with

respect to at least one contention will not be permitted to participate

as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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 30, 1998.

Description of amendments request: Currently, the Calvert Cliffs

Technical Specifications allow defective tubes to be plugged and

removed from service, or to be repaired by either the laser-welded

sleeving technique developed by Westinghouse Electric Corporation or by

using leak-tight, tungsten inert gas-welded sleeving developed by

Combustion Engineering, Inc. (ABB-CE). The proposed amendment will

revise the appropriate Technical Specifications to permit the use of

leak-limiting Alloy 800 repair sleeves developed by ABB-CE to be used

at Calvert Cliffs. Combustion Engineering provides two types of leak-

limiting Alloy 800 repair sleeves. The first type of repair sleeve

spans the expansion transition zone of the tube at the top of the

tubesheet. The second type of repair sleeve spans the degraded areas at

an eggcrate support elevation or in a free span section.

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 would not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The ABB CE Alloy 800 leak-limiting repair sleeves are designed

using the applicable American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code and, therefore, meet the design

objectives of the original steam generator tubing. The applied stresses

and fatigue usage for the repair sleeves are bounded by the limits

established in the ASME Code. Mechanical testing has shown that the

structural strength of repair sleeves under normal, upset, and faulted

conditions provides margin to the acceptance limits. These acceptance

limits bound the most limiting (three times normal operating pressure

differential) burst margin recommended by Regulatory Guide 1.121. Burst

testing of sleeved tubes has demonstrated that no unacceptable levels

of primary-to-secondary leakage are expected during any plant

condition.

The Alloy 800 repair sleeve Technical Specification depth-based

plugging limit is determined using the guidance of Regulatory Guide

1.121 and the pressure stress equation of ASME Code, Section III. A

bounding tube wall degradation growth rate per cycle and a

nondestructive examination uncertainty has been assumed for determining

the repair sleeve plugging limit.

Evaluation of the repaired steam generator tubes indicates no

detrimental effects on the sleeve or sleeve-tube assembly from reactor

system flow, primary or secondary coolant chemistries, thermal

conditions or transients, or pressure conditions as may be experienced

at Calvert Cliffs. Corrosion testing of sleeve-tube assemblies

indicates no evidence of sleeve or tube corrosion considered

detrimental under anticipated service conditions.

The implementation of the proposed amendment has no significant

effect on either the configuration of the plant, or the manner in which

it is operated. The consequences of a hypothetical failure of the

sleeved tube is bounded by the current steam generator tube rupture

analysis described in Calvert Cliffs Updated Final Safety Analysis

Report, Section 14.15. Due to the slight reduction in diameter caused

by the sleeve wall thickness, primary coolant release rates would be

slightly less than assumed for the steam generator tube rupture

analysis and, therefore, would result in lower total primary fluid mass

release to the secondary system. A main steam line break or feed line

break will not cause a SGTR [steam generator tube rupture] since the

sleeves are analyzed for a maximum accident differential pressure

greater than that predicted in the Calvert Cliffs safety analysis. The

minimal repair sleeve leakage that could occur during plant operation

is well within the Technical Specification leakage limits.

Therefore, BGE has concluded that the proposed change 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 kind of

accident from any other accident previously evaluated.

As discussed above, the Alloy 800 repair sleeves are designed using

the applicable ASME Code as guidance; therefore, it meets the

objectives of the original steam generator tubing. As a result, the

functions of the steam generators will not be significantly affected by

the installation of the proposed sleeve. The proposed repair sleeves do

not interact with any other plant systems. Any accident as a result

[[Page 2245]]

of potential tube or sleeve degradation in the repaired portion of the

tube is bounded by the existing tube rupture accident analysis. The

continued integrity of the installed sleeve is periodically verified by

the Technical Specification requirements.

The implementation of the proposed amendment has no significant

effect on either the configuration of the plant, or the manner in which

it is operated. Therefore, BGE [Baltimore Gas and Electric Company]

concludes that this proposed change does not create the possibility of

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

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

The repair of degraded steam generator tubes with Alloy 80 leak-

limiting repair sleeves restores the structural integrity of the

degraded tube under normal operating and postulated-accident

conditions. The design safety factors utilized for the repair sleeves

are consistent with the safety factors in the ASME Boiler and Pressure

Vessel Code used in the original steam generator design. The portions

of the installed sleeve assembly that represent the reactor coolant

pressure boundary can be monitored for the initiation and progression

of sleeve/tube wall degradation. Use of the previously identified

design criteria and design verification testing assures that the margin

to safety is not significantly different from the original steam

generator tubes.

Therefore, BGE concludes that the proposed change does not involve

a significant reduction in a margin of safety.

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

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

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

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.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: November 9, 1998.

Description of amendment request: The proposed amendments would

revise Technical Specification Table 3.3.3-2, ``Emergency Core Cooling

System Actuation Instrumentation Setpoints'' to modify the degraded

voltage second level undervoltage relay setpoint and allowable value.

This change was submitted in response to a concern identified during an

Electrical Distribution System Functional Inspection.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

probability or consequences of an accident previously evaluated?

The setpoint change does not change the logic or function of the

degraded voltage protection circuits as described in UFSAR [Updated

Final Safety Analysis Report] Section 8.2.3. They also do not reduce

the reliability of these circuits. The increase in the degraded voltage

protection circuit setpoint is conservative compared to the existing

setpoint. There is no change as a result of this amendment to the

underlying accident and transient analyses that support operations of

LaSalle County Station. Inadvertent or spurious operation of the

degraded voltage protection function will initiate loading of the safe

shutdown loads on the diesel generators and is not assumed to initiate

an accident. The proposed degraded voltage setpoints are low enough to

prevent spurious actuations given the expected offsite grid voltages.

After implementation of this amendment, no operator actions are

required for equipment operations in response to degraded voltage

conditions.

This change does not affect the initiators or precursors of any

accident previously evaluated. This change will not increase the

likelihood that a transient initiating event will occur because

transients are initiated by equipment malfunction and/or catastrophic

system failure.

The consequences of accidents previously evaluated are not

increased. The proposed change does not affect the required level of

availability of systems required to mitigate the accidents considered

in the analyses. The proposed changes will ensure that the Class 1E

equipment will be capable of starting and operating during a design

basis accident with degraded offsite grid voltage. The increase in the

level of confidence is the result of more rigorous methodology used to

determine limiting Class 1E bus voltages at the minimum expected

offsite AC voltage. These calculations demonstrate that the degraded

voltage relays will not actuate following a block start of the

electrical loads that are automatically actuated by or as a consequence

of the LOCA [loss-of-coolant accident] signal if the switchyard voltage

remains above 352 kV.

If the grid voltage drops below 352 kV, then the analytical limit

of 3814 volts for proper operation of class 1E loads connected to each

4.16 kV Class 1E bus is assured by transfer to the respective onsite

power sources (Emergency Diesel Generators (EDGs)) by the degraded

voltage logic.

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:

Setpoint methodology established the bases to ensure that, with

known errors, the relays will detect degraded voltage conditions and

transfer safety loads to the EDGs at a voltage level adequate to ensure

proper safety equipment performance and to prevent equipment damage.

The greater than or equal to 3870 volt setpoint and the greater

than or equal to 3814 volt allowable value includes adequate tolerance

to calibrate the relay trip units while ensuring that the Class 1E bus

voltage will remain above the analytical limits.

These setpoint changes will ensure that adequate voltages will be

available for the continuous operation of safety-related equipment

required to function during a LOCA. These proposed changes will also

ensure that adequate voltages will be available for starting any Class

1E equipment.

The proposed degraded voltage setpoint change does not change the

design of the degraded voltage protection system or its function to

protect against degraded offsite power. Actuation of the degraded

voltage protection system will initiate a sequence of events that will

start the EDG for the associated Class 1E bus, strip loads from the

Class 1E bus, open all feed breakers to the Class 1E bus, close the

Emergency feed breaker (thus energizing the Class 1E bus from the

respective EDG), and initiate starting of the Safe Shutdown equipment

supplied by the Class 1E bus.

Since the scope of this change does not affect the operation of the

auxiliary power system or any actions necessary to mitigate the

consequences of accidents or achieve safe shutdown, the

[[Page 2246]]

change does not involve a new or different accident scenario.

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 proposed amendment will allow the degraded voltage setpoint to

be conservatively established based on new engineering calculations

which consider the lowest expected offsite grid voltage and operation

of required Class 1E equipment under design basis accident loading

conditions.

The proposed degraded voltage setpoints will ensure that adequate

Class 1E bus voltage will be available to support starting and

operation of the required Class 1E loads. The proposed setpoint

includes instrument error to ensure that the lowest possible voltage

will not be lower than the degraded voltage analytical limits.

Additionally, the proposed setpoints are low enough to prevent spurious

actuations due to expected fluctuations in the grid voltage. The new

setpoints are also set with margin to the minimum Class 1E bus voltage,

which is based on a minimum grid voltage of 352 kV, which is less than

the expected grid voltage of 354 kV. The proposed changes will provide

an increase in the level of protection that currently exists and will

ensure the margin of safety is adequately maintained.

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 proposes to determine that the

requested amendments involve 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.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: November 30, 1998.

Description of amendment request: The Safe Shutdown Makeup Pump

(SSMP) allowed outage time (AOT) is being decreased from 67 days to 14

days.

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:

Does the change involve a significant increase in the probability

or consequences of an accident previously evaluated?

The change does not involve a significant increase in the

probability or consequences of an accident previously evaluated. The

proposed change to the Technical Specification Allowed Outage Time is

conservative with respect to current requirements. This change is being

proposed to establish an AOT for the SSMP that is equivalent to that

for the reactor core isolation cooling (RCIC) pump (14 day AOT) in

order to enhance system performance by assuring maximum SSMP pump

availability to a level consistent with RCIC. This is necessary since,

pursuant to Paragraph III.G.3 of 10 CFR 50, Appendix R, the SSMP is an

alternate system to the RCIC system. By ensuring equipment

availability, the probability or consequences of an accident previously

evaluated are not increased. In addition, the proposed change has no

impact on any accident initiators or initial condition assumptions for

accident scenarios. Onsite or offsite dose consequences resulting from

an event previously evaluated are not affected by this proposed

amendment request.

Therefore, this proposed amendment does not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

Does the change create the possibility of a new or different kind

of accident from any accident previously evaluated?

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

different kind of accident from any accident previously evaluated. The

proposed license amendment provides a reduction to a Technical

Specification Allowed Outage Time to enhance system performance by

assuring maximum SSMP pump availability to a level consistent with

RCIC. The proposed change is conservative with respect to the current

requirements. The proposed amendment does not involve any plant

physical changes that would create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Does the change involve a significant reduction in a margin of

safety?

The proposed change does not involve a significant reduction in a

margin of safety. The proposed change enhances system performance by

assuring maximum SSMP pump availability to a level consistent with

RCIC. Since this is a conservative change that will enhance the

performance of the SSMP system, it does not involve a significant

reduction in the margin of safety.

Therefore, this change does not involve a significant reduction in

a margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

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

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

NRC Project Director: Stuart A. Richards.

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

Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida

Date of amendment request: October 30, 1998.

Description of amendment request: The proposed amendment would

change the Crystal River Unit 3 (CR-3) Improved Technical

Specifications (ITS) to delete a note regarding the number of required

channels for the Degrees of Subcooling function, and to subdivide the

Core Exit Temperature (Backup) function into two new functions in ITS

Table 3.3.17-1, Post-Accident Monitoring Instrumentation.

These proposed ITS changes support modifications scheduled for

Refueling Outage 11 at CR-3. These modifications are intended to

significantly improve the reliability and availability of information

to the control room operators for verifying adequate core cooling is

maintained following a design basis accident. The proposed ITS change

deletes the note describing the use of the SPDS as a backup since the

SPDS will be the primary indication of subcooling margin after the

planned modifications are implemented.

[[Page 2247]]

The planned modifications will separate the sixteen core exit

thermocouples into two separate channels of eight core exit

thermocouples each. Following the modifications, there will be two core

exit thermocouples per channel located in each core quadrant. Each

separate channel of eight core exit thermocouples will have an

associated core exit temperature recorder on the main control board,

instead of the current three recorders, and will provide input into the

associated channel of SPDS for calculation of subcooling margin.

The proposed ITS change will subdivide the current Core Exit

Temperature (Backup) function into two new functions, Core Exit

Temperature (Thermocouple) function and Core Exit Temperature

(Recorder) function. For the Core Exit Temperature (Thermocouple)

function, the proposed ITS will require at least two OPERABLE core exit

thermocouples per core quadrant (at least one per channel) to provide a

representative distribution of temperatures across the core to the

operator. For the Core Exit Temperature (Recorder) function, both core

exit temperature recorders will be required OPERABLE.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

The proposed changes to the Degrees of Subcooling, Core Exit

Temperature (Thermocouple), and Core Exit Temperature (Recorder)

functions in the CR-3 Improved Technical Specifications (ITS) ensure

appropriate post-accident monitoring instrumentation is available for

use by the operators during implementation of emergency operating

procedures. These emergency operating procedures provide direction to

the operators for ensuring that actions required to mitigate the

effects of the previously evaluated design basis accidents are

performed. The instrumentation is used for monitoring by the operators

after an accident occurs, perform no automatic functions, and there are

no credible failures of this instrumentation which could initiate any

accident previously evaluated. Therefore, the probability of occurrence

of any accident previously evaluated is unaffected.

The availability and use of this instrumentation ensures that the

prescribed manual operator actions for mitigating the consequences of

an accident will be implemented when necessary, and that the operator

has sufficient information to verify required automatic actions have

occurred when necessary. The availability and use of the

instrumentation provides assurance that the consequences of accidents

will not be greater than that previously evaluated. The associated

modifications that are planned for these post-accident monitoring

instruments will enhance the reliability of the required indications to

the operators.

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

from previously evaluated accidents?

The proposed changes to this post-accident monitoring

instrumentation will ensure appropriate instrumentation is available

for use by the operators following a design basis accident. This

instrumentation is necessary for performing certain manual actions, or

to verify automatic actions have occurred, which are required to

mitigate the effects of a design basis accident. The instrumentation is

used for monitoring by the operators after an accident occurs, perform

no automatic functions, and there are no credible failures of this

instrumentation which could initiate a new or different kind of

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

accident occurring as a result of this passive instrumentation is not

created.

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

The proposed changes to this post-accident monitoring

instrumentation provide additional assurance that adequate

instrumentation is available for use by the operators to perform manual

actions, and to verify that automatic actions that are required to

mitigate the effects of a design basis accident have occurred. The

instrumentation is used for monitoring by the operators after an

accident occurs, and perform no automatic functions. The availability

and use of this instrumentation ensures that the prescribed manual

operator actions for mitigating the consequences of an accident will be

implemented when necessary, and that the operators have sufficient

information to verify required automatic actions have occurred when

necessary. These required manual and automatic actions are necessary to

preserve the margin of safety as defined in the CR-3 ITS. The

availability and use of this instrumentation provides assurance that

the existing margin of safety will be maintained, and assumptions

related to the margin of safety during mitigation of design basis

accidents will be preserved. Therefore, the existing 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 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: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC--A5A, P.O. Box 14042, St. Petersburg, Florida

33733-4042.

NRC Project Director: Frederick J. Hebdon.

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

Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida

Date of amendment request: November 24, 1998.

Description of amendment request: The proposed amendment would

change the CR-3 Improved Technical Specifications (ITS) in support of a

modification to install a diesel-driven emergency feedwater (EFW) pump

(EFP-3) which is intended to resolve capacity limitations of the CR-3

1A Emergency Diesel Generator (EGDG). The licensee has determined that

installation of EFP-3 involves an unreviewed safety question and also

requires changes and additions to the ITS and Bases.

EFP-3 will be installed as a functional replacement for EFP-1, the

motor-driven EFW pump. EFP-3 will start and provide controlled and

monitored EFW flow to both steam generators through the same EFW block

and control valves as EFP-1 currently uses. The licensee stated that

removing the auto-start logic from EFP-1 would eliminate the need to

perform EGDG-1A load management to accommodate emergency safeguards

(ES) loads required to mitigate design basis accidents. EFP-1 will

remain available as a manually started pump. The installation of EFP-3

will also permit other changes in system operation which are intended

to reduce reliance on operator actions to perform EGDG load management.

The proposed ITS and Bases changes fall into two categories: (1)

new or revised ITS and Bases to account for equipment changes

associated with the new EFP-3, and (2) those ITS and Bases requirements

being deleted because they were approved until Cycle 12 only.

[[Page 2248]]

The new ITS requirements and revisions (category 1 changes) involve

revised surveillance requirements (SR) and Bases for EFP-3 (SR 3.7.5),

and new ITS and Bases for the diesel fuel oil supply, lube oil and

starting air for EFP-3 (3.7.19). Also, the option to use EFP-3 for Once

Through Steam Generator (OTSG) cooling is added to the Bases for 3.4.6,

RCS Loops--MODE 5, Loops Filled. The Bases for 3.4.6, Background, lists

all feedwater pumps that may be available in MODE 5. EFP-3 is added

here for completeness.

The Bases for 3.7.5 are revised to describe the new EFP-3 and the

new role for EFP-1 as a manual defense-in-depth pump. The Bases are

also revised to indicate that EFP-3 cannot directly access the

condenser hotwell. The phrase ``with the exception of the loss of all

AC power (Ref. 3)'' is deleted from the Applicable Safety Analysis

because with the addition of EFP-3, the EFW system is able to maintain

its function on a loss of off-site power (LOOP) with a single failure.

In Section 3.7.5, EFW System, one SR is being revised and one new

SR is being added. These changes are intended to provide SRs that

demonstrate OPERABILITY of EFP-3 and essential subsystems. SR 3.7.5.1

and Bases are revised to add verification of proper valve position for

starting air and fuel oil flow paths for EFP-3 on a 45-day frequency.

SR 3.7.5.6 is added to provide assurance that the DC electrical

support system will be available to support OPERABILITY of EFP-3. This

SR is based on a similar SR currently approved for the station DC

system required by ITS 3.8.4, DC Sources--Operating. SR 3.7.5.6 was

determined necessary because DC power is essential for starting EFP-3.

ITS 3.7.19 was added to ensure essential subsystems are within

limits needed to maintain EFP-3 OPERABLE. The specification includes

requirements for fuel oil, lube oil, and starting air. This

specification has an allowed outage time (AOT) for these parameters if

they are less than the limit but above a minimum value. Below the

minimum allowed value, EFP-3 must be declared inoperable.

A number of ITS and Bases are being revised to remove the

requirements that permitted operation of CR-3 until Cycle 12 only

(category 2 changes). All text marked with the footnote ``Note--Valid

until Cycle 12 only,'' and the note itself, is being deleted except for

a few instances which are discussed in the licensee's submittal.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

4. Does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

This change involves the addition of a new safety-related Diesel-

Driven Emergency Feedwater Pump (EFP-3). The Emergency Feedwater (EFW)

System is not an initiator for any design basis accident except for

those accidents associated with an increase in primary to secondary

cooling and a loss of heat sink. The new EFP-3 functionally replaces

the Motor-Driven Emergency Feedwater Pump (EFP-1) and is no more likely

to cause an inadvertent cooldown than the existing EFP-1. The starting

logic of the Emergency Feedwater Initiation and Control system is the

same for EFP-3 as it was for EFP-1 before. No other control or logic

changes are being made that would make EFP-3 more likely to cause a

cooldown transient.

EFP-3 has a slightly greater probability of failing to start

compared to EFP-1 with offsite power available. Therefore, there is a

slight increase in the probability of an event that involves a loss of

heat sink when considering only the Improved Technical Specifications

(ITS) required EFW Pumps. The new EFP-3 will be highly reliable and

therefore this increase in risk is not significant. Loss of EFP-3 alone

does not cause a total loss of heat sink without the loss of the

Turbine-Driven Emergency Feedwater Pump (EFP-2) and the remaining

feedwater pumps. The most important of these feedwater pumps is EFP-1,

which will be maintained as a safety-grade backup. EFP-3 is less

reliable than EFP-1 with offsite power available. However, if offsite

power is not lost, EFP-1 should be available for use. Therefore, the

overall EFW system reliability is enhanced.

The consequences of the failure of EFP-3 to start or inadvertently

actuate were considered. Failure of EFP-3 to start will have the same

impact as failure of EFP-1. Therefore, the consequences of evaluated

accidents are the same. EFP-3 will be designed to have minimum and

maximum flows equivalent to EFP-1. No changes to the system will cause

a decrease in the ability of the EFW system to remove heat from the

Once Through Steam Generators (OTSGs). Similarly, the heat removal

capability of EFP-3 will not be different than EFP-1. Therefore, there

will not be the potential of a significantly greater overcooling event

due to inadvertent start of EFP-3.

The license changes associated with the addition of EFP-3 remove a

number of ITS Actions that established compensatory measures due to the

possibility of overloading the Emergency Diesel Generators (EGDGs) and

cross-train dependencies with EFP-2. These compensatory actions are no

longer required. The changes to the EFW system eliminate EGDG

limitations and reliance of the ``A'' train EFW pump on EFP-2. The

revised ITS Actions ensure the equipment required to mitigate an

accident is restored to OPERABLE status in accordance with previously

approved limits. In addition, replacing required operator actions with

automatic functions provides greater assurance that mitigating actions

will occur. Therefore, these changes will not adversely affect the

probability or consequences of evaluated accidents.

Based on the above, the addition of EFP-3 and the associated

license changes do not involve a significant increase in the

probability or consequences of a previously evaluated accident.

5. Does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

EFP-3 performs the same functions as the existing EFP-1. No plant

conditions are changed to cause new or different accidents. Although a

diesel engine has different failure modes than a motor-driven pump, the

consequences of a pump failure are the same. An interlock and

administrative controls are provided to ensure that both EFP-3 and EFP-

1 do not run at the same time. The interlock and administrative

controls prevent any new interactive failure modes that could be caused

by having both ``A'' train pumps (or all three EFW pumps) operating at

the same time.

The revised ITS Actions ensure equipment is restored to OPERABLE

status in accordance with previously approved timeframes. No new plant

configurations or conditions are created by these Actions.

Therefore, these changes cannot create the possibility of an

accident of a different type than previously evaluated in the SAR

[Safety Analysis Report].

6. Does not involve a significant reduction in the margin of

safety.

EFP-3 is designed to meet the same performance criteria as EFP-1.

EFP-3 will replace EFP-1 in the ITS. The pump will perform the same

functions, will be reliable and meet the same design criteria. There

are no functions performed by EFP-1 that will be significantly

different with EFP-3. The

[[Page 2249]]

margin of safety provided by the specification relates to the ability

to provide a heat sink. EFP-3 will provide the same margin of safety.

In addition, EFP-1 will be available as a safety-grade backup and can

deliver EFW to the OTSGs if offsite power is available or if the ``A''

train EGDG has adequate load margin.

The cooling capability of EFP-3 will be equivalent to EFP-1.

Therefore, EFP-3 provides the same protection to the fuel cladding from

temperature excursions as EFP-1. The EFP-3 modifications will be done

without making penetrations through reactor coolant system (RCS) or

containment boundaries. Therefore, the integrity of these fission

product barriers remains unchanged.

The proposed changes to the ITS delete temporary restrictions

placed on systems due to the potential to overload the EGDGs and cross-

train dependencies with EFP-2. These compensatory actions are no longer

required. The changes to the EFW system eliminate EGDG limitations and

reliance of the ``A'' train EFW pump on EFP-2. The revised ITS Actions

ensure the equipment required to mitigate an accident is restored to

OPERABLE status in accordance with previously approved limits. In

addition, replacing required operator actions with automatic functions

provides greater assurance that mitigating actions will occur.

Based on the above evaluation, there is no reduction in the margin

of safety associated with the proposed equipment, system and license

changes.

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: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida

33733-4042.

NRC Project Director: Frederick J. Hebdon.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of amendment request: July 14, 1998.

Description of amendment request: The proposed amendment would

amend the Technical Specifications to revise the liquid and gaseous

release rates to reflect the replacement of the former 10 CFR 20.106

requirements with the existing 10 CFR 20.1302 requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not:

1. Involve a significant increase in the probability or consequence

of an accident previously evaluated.

The likelihood that an accident will occur is neither increased nor

decreased by these Technical Specification changes. These Technical

Specifications changes will not impact the function or method of

operation of plant equipment. No systems, equipment, or components are

affected by the proposed changes. The proposed revisions to the liquid

and gaseous release rate limits will not result in any change or

increase in the types or amounts of effluents other than that which has

historically been deemed acceptable for release, nor will there be an

increase in individual or cumulative occupational radiation exposures

other than that which has historically been deemed acceptable.

Therefore, the proposed changes to the Technical Specifications do not

involve any increase in the probability or consequences of any accident

previously evaluated.

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

from any accident previously evaluated.

The proposed changes do not involve changes to the physical plant

or operations. The proposed changes are administrative in nature and

will not change the types and amounts of effluents from that which has

historically been deemed acceptable. Since these administrative changes

do not contribute to accident initiation, they do not produce a new

accident scenario nor do they alter any existing accident scenarios.

Therefore, the proposed changes to the Technical Specifications would

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

any accident previously evaluated.

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

The proposed changes will not reduce the margin of safety because

compliance with the limits of the existing 10 CFR 20.1301 will be

demonstrated by operating within the limits of 10 CFR Part 50, Appendix

I and 40 CFR Part 190. For the liquid effluent releases the annual dose

of 500 mrem, upon which the concentrations in the previous 10 CFR Part

20, Appendix B, Table II, Column 2, are based, is a factor of 10 higher

than the annual dose of 50 mrem, upon which the concentrations in the

existing 10 CFR 20, Appendix B, Table 2, Column 2, are based. Also, for

gaseous effluent releases, the limits associated with the gaseous

release Technical Specifications will be revised to the previously

acceptable instantaneous dose rate limits.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME 04578.

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 321 Ferry Road, Wiscasset, ME 04578.

NRC Project Director: Seymour H. Weiss.

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

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

Connecticut

Date of amendment request: December 10, 1998.

Description of amendment request: The proposed amendment would

allow NNECO to implement plant modifications that would ensure that

proper flow paths can be established for boron precipitation control

after a loss-of-coolant accident.

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 10 CFR 50.92, 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 10 CFR 50.92(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 plant modifications will ensure proper flow paths can

be established for boron precipitation control after a Loss of Coolant

Accident

[[Page 2250]]

(LOCA). This will be accomplished by the following plant modifications:

a. Provide an alternate AC source of power for 2-SI-651, ``Shutdown

Cooling Header Containment Isolation Valve,'' a Facility Z1 component,

from Facility Z2.

b. Provide an alternate DC source of power for 2-CH-517,

``Auxiliary Spray Charging Header Supply Valve,'' a Facility Z2

component, from Facility Z1.

c. Provide an alternate DC source of power for 2-CH-519, ``Loop 1A

Charging Header Supply Valve,'' a Facility Z2 component, from Facility

Z1.

d. Provide test jacks to determine valve position for LPSI [low-

pressure safety injection] injection valves 2-SI-615, 2-SI-625, 2-SI-

635, and 2-SI-645 at the respective motor control center (MCC [motor

control center] B51 for 2-SI-615 and 2-SI-625 (MCC B61 for 2-SI-635 and

2-SI-645).

e. Provide bypass capability of the low pressure open permissive

for 2-SI-651.

The alternate power supply to valves 2-SI-651, 2-CH-517 and 2-CH-

519, and the position indication for valves 2-SI-615, 2-SI-625, 2-SI-

635, and 2-SI-645 cannot initiate an accident. The proposed

modifications will not change the design parameters, failure positions

or design requirements of the valves. The proposed plant modifications

will ensure valves 2-SI-651, 2-CH-517 and 2-CH-519 can operate after a

LOCA to perform their accident mitigating functions. Therefore,

providing an additional power source to 2-SI-651, 2-CH-517 and 2-CH-

519, and a local means of determining the position of valves 2-SI-615,

2-SI-625, 2-SI-635, and 2-SI-645 cannot initiate an accident and will

not adversely affect the function of these components to mitigate the

consequences of an accident.

The proposed plant modifications will also bypass the open

permissive for 2-SI-651. This pressure permissive, which protects the

low pressure Shutdown Cooling (SDC) System from the high pressure

Reactor Coolant System (RCS), allows 2-SI-651 to be opened only when

pressurizer pressure is below 280 psia [pounds per square inch

absolute]. This pressure permissive would be disabled upon a loss of

Facility Z1 power. This would prevent the opening of 2-SI-651. The new

local control switch for 2-SI-651, which bypasses this pressurizer

pressure permissive, is isolated by normally open relay contacts. When

aligned to its alternate power, local control is enabled, and remote

control in the Main Control Room is isolated. Multiple operator errors

would be required to align 2-SI-651 to the alternate power source

during normal operation. To misalign these valves, an equipment

operator would have to perform steps located only in an Emergency

Operating Procedure. Additionally, control room operators would have to

disregard annunciators that indicate the valves are being transferred

to their alternate power source. Therefore, the only time the valve is

expected to be opened by the local control switch is after a LOCA with

a Facility Z1 failure. Currently, the potential exists for an operator

to open the valve when pressure is above 280 psia. An undetectable

single failure of the contact which provides the permissive would allow

an operator to open the valve even when pressure is above 280 psia.

During normal operation, this condition would be annunciated in the

Main Control Room. During accident conditions, this annunciator may be

disabled. Therefore, 2-SI-651 could be opened with pressurizer pressure

above 280 psia without annunciation. Although 2-SI-651 could be opened,

the pressure permissive for 2-SI-652, the upstream isolation valve

(Attachment 1 Figure 1), would prevent 2-SI-652 from opening. This

would protect the shutdown cooling suction line from

overpressurization. During accident conditions, if both valves were

opened and pressure increased above 280 psia, annunciation of 2-SI-652

being open would be available to provide indication of the potential

overpressure condition. Therefore, the installation of the capability

to bypass the open permissive for 2-SI-651 will not result in a

significant increase in the probability or consequences of an accident

previously evaluated.

The proposed plant modifications have no adverse effect on how any

of the associated systems or components function to prevent or mitigate

the consequences of design basis accidents. Also, the proposed changes

have no adverse effect on any design basis accident previously

evaluated since the modifications will ensure that accident mitigation

equipment will be available to function as assumed in the LOCA

analysis. Therefore, the proposed plant modifications do not result in

a significant increase in the probability or consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

The proposed plant modifications will provide the capability of

powering 2-SI-651, 2-CH-517 and 2-CH-519 from either Facility Z1 or

Facility Z2, and will add test jacks to MCC B51 and B61 to determine

the position of valves 2-SI-615, 2-SI-625, 2-SI-635, and 2-SI-645.

Additionally, this activity adds a local control switch which will

bypass the open permissive for 2-SI-651 when aligned to the alternate

power source. A single failure in any of the breakers or disconnect

switches which allow 2-SI-651, 2-CH-517 and 2-CH-519 to be powered from

either facility is bounded by the failure of the valve. A failure of

any of the test jacks may result in a loss of control power to the

associated valves. This failure is also bounded by the failure of the

valve. During normal operation the local control switch which bypasses

the pressure permissive is isolated by normally open contacts. A single

failure of the local control switch or isolating relay during normal

operation cannot disable the pressure permissive.

Since a single failure of any component added by this activity is

bounded by existing component failures, a failure of these components

cannot create a new accident. Therefore, the proposed plant

modifications will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

The proposed plant modifications will ensure boron precipitation

control can be established. This will be accomplished by providing an

alternate power source for 2-SI-651, 2-CH-517 and 2-CH-519, and adding

test jacks to determine the position of valves 2-SI-615, 2-SI-625, 2-

SI-635, and 2-SI-645. Additionally, this activity adds a local control

switch which will bypass the open permissive for 2-SI-651 when aligned

to the alternate power source. Although the potential exists to route

redundant power trains in the same cable trays, conduits and cable, the

design of the modifications ensures that a single failure will not

compromise the redundant power distribution system. The installation of

the connection jacks and local control switch will not alter the

failure analysis for the valves, and will not change the design

parameters of the valves (i.e. pressure rating). Therefore, the

proposed plant modifications will not compromise RCS pressure

boundaries, containment integrity, or fuel cladding. In addition, the

new disconnect switches, breakers, cabling, and auxiliary components

are all designed for the rated voltages and currents, and are QA

[quality assurance] Category I seismically and environmentally

qualified, as required.

[[Page 2251]]

Based on the above, the proposed plant modifications will not

reduce the integrity of the plant protective boundaries, or adversely

affect the LOCA analysis. These modifications will have no adverse

effect on equipment important to safety. The equipment will continue to

function as assumed in the design basis accident analysis. This will

ensure that the acceptance criteria of 10 CFR 50.46(b)(5) for long term

core cooling will be met. Therefore, there will be no significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Project Director: William M. Dean.

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

Generating Station, Salem County, New Jersey

Date of amendment request: December 16, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Surveillance Requirements 4.8.1.1.2

and 4.8.1.1.3, Table 4.8.1.1.2-1, and the associated Bases. The

proposed changes would remove the Emergency Diesel Generator

accelerated testing and special reporting requirements from the TSs in

accordance with the guidance provided in Generic Letter 94-01.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed TS changes do not involve any physical changes to

plant structures, systems, or components. [Public Service Electric &

Gas Company] PSE&G has implemented the provisions of the Maintenance

Rule for diesel generators, including the associated regulatory

guidance, thereby establishing a program that assures diesel generator

performance. The elements of the program include the performance of

detailed root cause analysis of individual failures, effective

corrective actions taken in response to individual failures, and

implementation of preventive maintenance consistent with the

Maintenance Rule. Monitoring the effectiveness of diesel generator

maintenance and continuing surveillance testing in accordance with the

proposed TS changes will ensure that the diesel generators will perform

their intended functions and will minimize failures. The accelerated

testing requirements are therefore no longer considered to be necessary

and are deleted. The requirements of 10 CFR 50.72 and 10 CFR 50.73

ensure that diesel generator failures are properly reported. The

special reporting requirements are therefore unnecessary and are

deleted. Based on the above information, the changes will not adversely

affect the assurance of diesel generator reliability or operability,

and there is no significant increase in the probability or consequences

of any accident previously evaluated.

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

different kind of accident from any accident previously analyzed.

The proposed TS changes do not involve any physical changes to the

design of plant systems, structures or components, nor do the changes

involve a change in plant operation. The diesel generators will

continue to function as designed to mitigate the consequences of an

accident. Eliminating the accelerated testing requirements and special

reporting requirements does not permit plant operation in a

configuration that would create a different type of malfunction to the

diesel generators than any previously evaluated. In addition, the

proposed TS changes do not alter the conclusions described in the

[Updated Final Safety Analysis Report] UFSAR regarding the safety

related functions of the diesel generators or their support systems. No

new failure modes will be introduced. Therefore, the proposed changes

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

from any accident previously evaluated.

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

a margin of safety.

This request does not involve an adverse impact on diesel generator

design, operation, or reliability. Since monitoring and maintenance is

being performed in conformance with 10 CFR 50.65, modifying the

surveillance testing frequency requirements does not adversely affect

the reliability of the diesel generators. Deletion of the special

reporting requirements does not impact operability or reliability of

the diesel generator. Since the diesel generator function is not

affected by the proposed change, this request does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

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

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

NRC Project Director: Robert A. Capra.

Previously Published Notices of Consideration of Issuance of

Amendments to Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

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: December 3, 1998.

Brief description of amendments: The amendments would revise

Technical Specification Section 4.6.5.1, ``Ice Condenser, Ice Bed,''

and the associated bases to reflect the maximum ice condenser flow

channel blockage assumed in the accident analyses.

[[Page 2252]]

Date of publication of individual notice in Federal Register:

December 28, 1998.

Expiration date of individual notice: January 27, 1999.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Notice of Issuance of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

50-529, Palo Verde Nuclear Generating Station, Units Nos. 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendment: October 6, 1998.

Brief description of amendment: The amendments revise Technical

Specifications (TS) 3.3.1, ``Reactor Protective System (RPS)

Instrumentation--Operating,'' and TS 3.3.2, ``Reactor Protective System

(RPS) Instrumentation--Shutdown.'' The amendments clarify the power

level threshold at which certain RPS instrumentation trips must be

enabled and may be bypassed, and clarify that this level is a

percentage of the neutron flux at rated thermal power (RTP). The bypass

power level, 1E-4% RTP, is specified as logarithmic power instead of

thermal power. The NRC approved these changes for Palo Verde Unit 3 on

an exigent basis in its letter dated October 19, 1998. The exigent TS

amendment resulted in TS pages with notes specifying different

requirements between Unit 3 and Units 1 and 2. These amendments remove

these notes regarding Unit 3 from the affected TS pages so that all

Units now have the same TS.

Date of issuance: December 23, 1998.

Effective date: December 23, 1998.

Amendment No.: Unit 1-119; Unit 2-119.

Facility Operating License Nos. NPF-41 and NPF-51: The amendment

revised the Technical Specifications.

Date of initial notice in Federal Register: November 4, 1998 (63 FR

59586).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of application for amendment: February 11, 1998.

Brief description of amendment: The Updated Final Analysis Report

(UFSAR) describes the response of the salt service water (SSW) system

to a complete loss of AC power by assuming that the system would be

divided by the closure of one of the two division isolation valves.

Boston Edison Company (BECo) has discovered single failures involving a

partial loss of AC power could place the SSW system in a configuration

of one pump supplying both trains of heat exchangers for the first 10

minutes of the worst case design basis accident. BECo has determined

that these single failures are an unreviewed safety question. The

amendment authorizes BECo to change UFSAR Section 10.7, ``Salt Service

Water System,'' to address this single fauilure vulnerability.

Date of issuance: December 21, 1998.

Effective date: December 21, 1998.

Amendment No.: 180.

Facility Operating License No. DPR-35: Amendment revised the UFSAR.

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

17220)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 21, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

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: March 10, 1997, as supplemented

May 23, 1997, and October 15, 1998.

Brief description of amendment: This amendment revises Technical

Specification (TS) 3.5.1, ``Emergency Core Cooling System (ECCS)

Accumulators,'' by (1) increasing the allowed outage time (from 1 hour

to 72 hours) that one ECCS accumulator can be inoperable as a result of

the boron concentration being outside of TS limits, and (2) modifying

surveillance requirement 4.5.1 consistent with the guidance provided in

NUREG-1366, ``Improvements to Technical Specifications Surveillance

Requirements,'' December 1992, and the Standard Technical

Specifications (STS) for Westinghouse Plants, NUREG-1431, Revision 1.

Date of issuance: December 31, 1998.

Effective date: December 31, 1998.

Amendment No.: 86.

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

Technical Specifications

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

17226).

The May 23, 1997, and October 15, 1998, submittals contained

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 December 31, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

[[Page 2253]]

Consumers Energy Company, Docket No. 50--155, Big Rock Point (BRP)

Plant, Charlevoix County, Michigan

Date of application for amendment: September 19, 1997.

Brief Description of amendment: This amendment changes the DPR--6

License and revises its Technical Specifications to reflect the

permanently shutdown and defueled condition of the BRP plant.

Date of issuance: December 24, 1998.

Effective date: No later than 45 days from date of issuance.

Amendment No.: 120.

Facility Operating License No. DPR-6: The amendment revises the

DPR--6 License and Appendix A Technical Specifications to the licensee.

Date of initial notice in Federal Register: December 3, 1997 (62 FR

63974).

No significant hazards consideration comments received: No.

Local Public Document Room location: North Central Michigan College

Library, 1515 Howard Street, Petoskey, MI 49770.

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

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

Date of application for amendments: July 13, 1998.

Brief description of amendments: These amendments change the Beaver

Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2) Updated

Final Safety Analysis Reports (UFSAR) descriptions of the Intake

Structure main entrance and interconnecting cubicle doors. The changes

approved by these amendments address a new failure mode of safety-

related equipment that had not been previously considered for BVPS-1.

The changes state that the cubicle interconnecting flood protection

doors are normally closed with their inflatable seals depressurized and

that the associated security/fire doors are normally closed. This door

closure arrangement provides protection for the safety-related

equipment in the interconnecting cubicles from the consequences of

potential internal flooding.

Date of issuance: December 16, 1998.

Effective date: December 16, 1998.

Amendment Nos.: 218 and 96.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

approve changes to the Updated Final Safety Analysis Reports.

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

43202)

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 16, 1998.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001.

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power

Station, Unit No. 1, Shippingport, Pennsylvania

Date of application for amendment: June 18, 1996, as supplemented

September 8 and 30, 1998.

Brief description of amendment: The amendment (1) makes editorial

changes to Technical Specification (TS) 4.4.5 and associated Bases; (2)

revises the Bases for TS 3.4.6.2 to provide consistency with the Beaver

Valley Power Station, Unit No. 1, Updated Final Safety Analysis Report

(UFSAR); and (3) revises Index Page XVII to reflect the revision of

page numbers due to shifting of text by License Amendment No. 198.

Date of issuance: December 21, 1998.

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment No: 219.

Facility Operating License No. DPR-66. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64109).

The September 8 and 30, 1998, letters did not change the initial

proposed no significant hazards consideration determination or expand

the amendment request beyond the scope of the November 18, 1998,

Federal Register notice; these letters only provided updated TS pages

to be consistent with the UFSAR.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 21, 1998.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

No. 1, Pope County, Arkansas

Date of amendment request: May 31, 1996.

Brief description of amendment: The amendment revises the

surveillance test interval for the reactor protection system reactor

trip breakers, reactor trip modules, and electronic trip relays from a

monthly interval to a quarterly interval.

Date of issuance: December 31, 1998.

Effective date: December 31, 1998.

Amendment No.: 194.

Facility Operating License No. DPR-51: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 28, 1996 (61 FR

44356).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 31, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

2, Pope County, Arkansas

Date of amendment request: May 18, 1998, as supplemented on

December 8, 1998.

Brief description of amendment: The amendment allows the use of

trisodium phosphate stored in three baskets on the containment floor as

a replacement to the sodium hydroxide addition system for the control

of sump pH during long term core cooling in recirculation phase.

Date of issuance: December 23, 1998.

Effective date: The license amendment is effective as of its date

of issuance to be implemented prior to the facility's restart from

refueling outage 2R13.

Amendment No.: 194.

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

Technical Specifications.

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

56241)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

No. 2, Pope County, Arkansas

Date of application for amendment: July 28, 1997.

Brief description of amendment: The amendment modifies the actions

associated with Technical Specification (TS) Table 3.3-1 for the

Reactor Protective Instrumentation and TS Table 3.3-3 for the

Engineered Safety Feature Actuation System Instrumentation.

Date of issuance: December 29, 1998.

Effective date: December 29, 1998, to be implemented within 30

days.

[[Page 2254]]

Amendment No.: 195.

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

Technical Specifications.

Date of initial notice in Federal Register: August 27, 1997 (62 FR

45456).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

No. 2, Pope County, Arkansas

Date of application for amendment: June 29, 1998, as supplemented

by letters dated December 17, 1998 and December 22, 1998.

Brief description of amendment: This amendment revises the as-found

lift setting tolerance for the ANO-2 main steam safety valves and the

pressurizer safety valves, revises the maximum allowable linear power

level-high trip setpoint with inoperable steam line safety valves, and

relocates part of the specifications for steam line safety valves to

the ANO-2 Safety Analysis Report. Administrative and bases changes have

also been made.

Date of issuance: December 31, 1998.

Effective date: The license amendment is effective as of its date

of issuance to be implemented within 30 days.

Amendment No.: 197.

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

Technical Specifications.

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

56242).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 31, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: August 12, 1998.

Brief description of amendment: The amendment changes the Technical

Specifications by increasing the maximum boron concentration in the

Safety Injection Tanks (SITs) and the Refueling Water Storage Pool

(RWSP) from 2300 ppm to 2900 ppm.

Date of issuance: December 21, 1998.

Effective date: December 21, 1998, to be implemented within 60

days.

Amendment No.: 147.

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

Technical Specifications.

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

56249).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 21, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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: June 11, 1996, and supplemented

March 26, 1997.

Brief description of amendments: The amendments relocate certain

quality assurance related requirements from the TS to the licensee's

Quality Assurance Program Description.

Date of issuance: December 28, 1998.

Effective date: December 28, 1998, with full implementation within

120 days.

Amendment Nos.: 226 and 210.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 31, 1996 (61 FR

40022).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 28, 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.

Indiana Michigan Power Company, Docket No. 50-315 , Donald C. Cook

Nuclear Plant, Unit 1, Berrien County, Michigan

Date of application for amendment: August 28, 1998, as supplemented

November 4, 1998.

Brief description of amendment: The amendment grants relief from

the steam generator surveillance requirement in Section 4.4.5.3 of the

Technical Specifications (TS). The surveillance requirement is

associated with non-destructive examination of the steam generator

tubes which is required every 24 months. The relief allows the

examination to be deferred from April 8, 1999, until the next refueling

outage for D.C. Cook, Unit 1.

Date of issuance: December 30, 1998.

Effective date: December 30, 1998, with full implementation within

45 days.

Amendment No.: 227.

Facility Operating License No. DPR-58: Amendment adds paragraph

2.C.(9) to the License.

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

53950).

The November 4, 1998, submittal provided additional information

that did not change the initial no significant hazards consideration

determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 30, 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. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of application for amendment: April 30, 1997, as supplemented

November 12, 1998.

Brief description of amendment: The amendment deletes TSs

requirements associated with meterological monitoring instrumentation

which have been relocated to the Updated Safety Analysis Report in

accordance with 10 CFR 50.36 and the guidance in NRC Generic Letter 95-

10, ``Relocation of Selected Technical Specification Requirements

Related to Instrumentation.''

Date of issuance: December 22, 1998.

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

30 days.

Amendment No.: 85.

Facility Operating License No. NPF-69: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33126).

The November 12, 1998, 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 December 22, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

[[Page 2255]]

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

York 13126.

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

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

Connecticut

Date of application for amendment: September 19, 1995.

Brief description of amendment: The amendment reduces the frequency

of the Technical Specification (TS) 4.5.1.d surveillance interval for

boron concentration of the safety injection tasks from once per 31 days

to once every 6 months. Initially, the change was requested for TS

Section 4.5.1.b. However, TS Section 4.5.1.b was subsequently changed

to TS Section 4.5.1.d by Amendment No. 220 to Facility Operating

License No. DPR-65 dated September 3, 1998, in response to NNECO's

application dated August 23, 1995.

Date of issuance: December 17, 1998.

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

60 days from the date of issuance.

Amendment No.: 221.

Facility Operating License No. DPR-65: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 25, 1995 (60 FR

54722).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 17, 1998

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Date of application for amendment: July 2, 1998.

Brief description of amendment: The amendment revises the Updated

Final Safety Analysis Report (UFSAR) by changing UFSAR Sections 9.7.2,

``Service Water,'' and 9.4, Reactor Building Closed Cooling Water,'' to

include in the discussions the use of various types of internal

protective coatings and liners used in the piping and components of the

systems. The change also indicates that periodic maintenance,

surveillance, and inspections will be conducted to ensure that coating

or liner degradation will be promptly detected and corrected to provide

reasonable assurance that the systems can perform their safety-related

functions.

Date of issuance: December 18, 1998.

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

60 days from the date of issuance.

Amendment No.: 222.

Facility Operating License No. DPR-65: Amendment revised the

Updated Final Safety Analysis Report and Appendix B to Operating

License.

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

43206).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 18, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of application for amendment: July 5, 1995, as supplemented

October 9, 1998.

Brief description of amendment: The amendment extends surveillance

test intervals and allowable out-of-service times for instrumentation

in the Emergency Core Cooling (ECCS), Rod Block, Isolation Group 4

(High Pressure Coolant Injection, or HPCI) and Isolation Group 5

(Reactor Core Isolation Cooling, or RCIC), Reactor Building Ventilation

& Standby Gas Treatment, Recirculation Pump Trip and Alternate Rod

Injection, and Shutdown Cooling Supply Isolation Systems.

Date of issuance: December 23, 1998.

Effective date: December 23, 1998, with full implementation within

30 days.

Amendment No.: 103.

Facility Operating License No. DPR-22. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 30, 1995 (60 FR

45182). The October 9, 1998, submittal withdrew a portion of the

original request, made additional editorial changes, and provided

updated Technical Specification pages. This information was within the

scope of the original Federal Register notice and did not change the

staff's initial proposed no significant hazards considerations

determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of application for amendment: August 15, 1996, as supplemented

March 19 and October 12, 1998.

Brief description of amendment: The amendment revises the Technical

Specifications so that either 8 or 12 hour shifts will be considered

``normal'' and 40 hours will be considered a ``nominal'' week, changes

the wording for surveillances required ``once per shift'' to ``once per

12 hours,'' clarifies the ``once per hour'' wording related to fire

watch patrols, and makes a number of other clarifications and

typographical corrections.

Date of issuance: December 24, 1998.

Effective date: December 24, 1998, with full implementation within

30 days.

Amendment No.: 104.

Facility Operating License No. DPR-22: Amendment revised the

Technical Specifications.

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

53951). The October 12, 1998, submittal provided additional

clarifications and new TS pages. This information was within the scope

of the original Federal Register notice and did not change the staff's

initial proposed no significant hazards considerations determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 24, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

[[Page 2256]]

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments: December 14, 1995, as

supplemented on November 25, 1996, April 10, September 4, and December

29, 1997, January 8, March 2, June 11, August 12, and October 30, 1998.

Brief description of amendments: The amendments revise Technical

Specifications (TS) Table of Contents; TS 3.1, ``Reactor Coolant

System;'' TS 4.0, ``Surveillance Requirements;'' TS 5.0, ``Design

Features;'' and associated Bases by removing or relocating requirements

that are adequately controlled by existing regulations other than 10

CFR 50.36 and the TS and by modifying TS 6.0 to more closely meet the

format and content of the standard technical specifications.

Date of issuance: December 7, 1998.

Effective date: December 7, 1998, with full implementation of the

TS and License Condition 7 by September 1, 1999. License Condition 6

shall be implemented by the next USAR update, but no later than June 1,

1999. Implementation shall also include the relocation of TS

requirements to the appropriate licensee-controlled documents as

identified in the licensee's application dated December 14, 1995, as

supplemented on November 25, 1996, April 10, September 4, and December

29, 1997, January 8, March 2, June 11, August 12, and October 30, 1998,

and evaluated in the staff's safety evaluation attached to these

amendments.

Amendment Nos.: 141 and 132.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments

revised the Licenses and TS.

Date of initial notice in Federal Register: June 5, 1996 (61 FR

28618). The November 25, 1996, April 10, September 4, and December 29,

1997, January 8, March 2, June 11, August 12, and October 30, 1998,

submittals provided additional clarifying information, revised

implementation dates, and updated TS pages. This information was within

the scope of the original Federal Register notice and did not change

the staff's initial proposed no significant hazards considerations

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 7, 1998. `

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments: September 4, 1998.

Brief description of amendments: The amendments revise Technical

Specifications (TS) 3.1.A.3.b, 4.18, and Bases for TS 4.18 to clarify

the surveillance requirements and limiting conditions for operation of

the reactor coolant vent system.

Date of issuance: December 17, 1998.

Effective date: December 17, 1998, with full implementation within

30 days.

Amendment Nos.: 142 and 133.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments

revised the TS.

Date of initial notice in Federal Register: September 23, 1998 (63

FR 50938).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 17, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,

Unit No. 1, Washington County, Nebraska

Date of amendment request: April 17, 1997.

Brief description of amendment: The amendment revises Technical

Specifications (TS) 2.12, ``Control Room Systems,'' to delete the

limiting condition for operation (LCO) and surveillance for control

room temperature and replace it with an associated LCO and surveillance

for the control room air conditioning system. In addition, the

amendment revises TS 2.1, ``Reactor Coolant System,'' TS 2.6,

``Containment System,'' and TS 2.8, ``Refueling Operations,'' and the

associated surveillance requirements to incorporate the design basis

requirements for refueling operations and to correspond to NUREG-1432,

``Standard Technical Specifications Combustion Engineering Plants.''

Date of issuance: December 31, 1998.

Effective date: December 31, 1998, to be implemented within 60 days

from the date of issuance.

Amendment No.: 188.

Facility Operating License No. DPR-40. Amendment revised the

Technical Specifications.

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

30639).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 31, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102.

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: April 16, 1998, as supplemented

August 20, 1998.

Brief description of amendment: The amendment will extend the

surveillance interval for five instrument channels from the current 18

months to 24 months. The proposed amendment also revises Section 6 of

the Technical Specifications to reflect updated analyses.

Date of issuance: December 16, 1998.

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

30 days.

Amendment No.: 185.

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

Technical Specifications.

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

56256).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 16, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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: July 6, 1998.

Brief description of amendment: The amendment revises Appendix B

Technical Specification 3.5, Main Condenser Steam Jet Air Ejector and

Table 3.10-1, Radiation Monitoring Systems that Initiate and /or

Isolate Systems including the associated Bases to provide Allowable

Outage Times for selected instrumentation.

[[Page 2257]]

Date of issuance: December 28, 1998.

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

60 days.

Amendment No.: 249.

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

Technical Specifications.

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

43211).

Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated December 28, 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.

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

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

Jersey

Date of application for amendments: August 1, 1997, as supplemented

on October 6, 1997, February 18 and July 7, 1998.

Brief description of amendments: The amendments revise Technical

Specification Section 4.2.1 of Appendix B to require that Public

Service Electric & Gas Company (PSE&G) adhere to the Incidental Take

Statement, approved by the National Marine Fisheries Service (NMFS),

but remove the specific requirements. Removing the specific

requirements of Section 4.2.1 enables PSE&G to utilize relief granted

by the NMFS on a case-by-case basis.

Date of issuance: December 18, 1998.

Effective date: Effective as of its date of issuance, to be

implemented within 60 days.

Amendment Nos: 216 and 196.

Facility Operating License Nos. DPR-70 and DPR-75: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 10, 1997 (62

FR 47698).

The October 6, 1997, February 18 and July 7, 1998 submittals

provided clarifying information that did not change the initial

proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 18, 1998.

No significant hazards consideration comments received: No.

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

West Broadway, Salem, NJ 08079.

South Carolina Electric & Gas Company, South Carolina Public Service

Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit

No. 1, Fairfield County, South Carolina

Date of application for amendment: May 21, 1997, as supplemented on

December 4, 1998. The December 4, 1998, submittal contained clarifying

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

consideration determination.

Brief description of amendment: The amendment revises the Virgil C.

Summer Nuclear Station Technical Specifications to change the methods

for testing the control room and spent fuel pool ventilation system

charcoal adsorbers from American National Standards Institute Standard

N509-1980 to American Society for Testing and Materials Standard D3803-

1989.

Date of issuance: December 23, 1998.

Effective date: December 23, 1998.

Amendment No.: 140.

Facility Operating License No. NPF-12: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33133).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

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

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

Diego County, California

Date of application for amendments: May 7, 1998.

Brief description of amendments: The amendments revise the

reference for obtaining the thyroid dose conversion factors used in the

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

Specification Section 1.1, ``Definitions.''

Date of issuance: December 16, 1998.

Effective date: December 16, 1998, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 2--145; Unit 3--137.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 4, 1998 (63 FR

59595).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 16, 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.

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

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

Diego County, California

Date of application for amendments: June 12, 1998, as supplemented

by letters dated September 18, 1998, October 29, 1998, and November 23,

1998.

Brief description of amendments: The amendments authorize revision

of the San Onofre Nuclear Generating Station Updated Final Safety

Analysis Report to incorporate a new turbine missile protection

calculation methodology.

Date of issuance: December 21, 1998.

Effective date: December 21, 1998, to be implemented in the next

periodic update of the UFSAR in accordance with 10 CFR 50.71(e) that

occurs after 60 days of the date of issuance.

Amendment Nos.: Unit 2-146; Unit 3-138.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

authorize revisions to the Updated Final Safety Analysis Report.

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

60412).

The November 23, 1998, supplemental letter provided additional

information and did not change the original no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 21, 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 No. 50-259, Browns Ferry Nuclear

Plant, Unit No. 1, Docket No. 50-260, Browns Ferry Nuclear Plant, Unit

No. 2, and Docket No. 50-296, Browns Ferry Nuclear Plant, Unit No. 3,

Limestone County, Alabama

Date of amendment request: June 2, 1997 as supplemented November

19, 1998.

Description of amendment request: Presently Technical Specification

(TSs) require both the recirculation loops to

[[Page 2258]]

be operable and provide a 12-hour allowable outage time (AOT) for

single loop operation (SLO) mode. The amendments modify TS to allow

indefinite SLO instead of the 12-hour AOT.

Date of issuance: December 23, 1998.

Effective date: December 23, 1998.

Amendment No.: 236, 256 and 216.

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

Amendments revised the Technical Specifications.

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

43377). The licensee's letter of November 19, 1998, did not expand the

scope of the application or affect the staff's initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611.

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

Unit 1, Rhea County, Tennessee

Date of application for amendment: February 18, 1998.

Brief description of amendment: Changes Technical Specification

(TS) 3.7.4, Steam Generator Atmospheric Dump Valves (ADVs), and its

associated bases by adding a new TS CONDITION, REQUIRED ACTION, and

COMPLETION TIME to address a potential condition where two ADVs are

made technically inoperable when one train of the safety-related

auxiliary control air system is taken out of service.

Date of issuance: December 17, 1998.

Effective date: December 17, 1998.

Amendment No.: 16.

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

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

43213) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 17, 1998.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

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

Unit 1, Rhea County, Tennessee

Date of application for amendment: February 28, 1996, as

supplemented October 2 and December 12, 1997, March 30 and December 11,

1998.

Brief description of amendment: The February 28, 1996 letter

proposed to extend the surveillance interval for Westinghouse type AR

relays with alternating current and direct current coils from quarterly

to an 18 month interval. The letter of December 11, 1998 revised the

scope of the application such that it now applies only to Westinghouse

type AR relays which use alternating current coils. Accordingly, this

amendment approves the extension of the surveillance interval only for

Westinghouse type AR relays which use alternating current coils.

Date of issuance: December 30, 1998.

Effective date: December 30, 1998.

Amendment No.: 17.

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

Technical Specifications.

Date of initial notice in Federal Register: April 10, 1996 (61 FR

15998). The October 2 and December 12, 1997, March 30 and December 11,

1998 letters provided clarifying information that did not change the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 30, 1998.

No significant hazards consideration comments received: No.

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: June 30, 1998, as supplemented

by submittals dated October 27, November 30, and December 3, 1998. The

supplemental submittals did not expand the scope of the original

application or change the staff's proposed no significant hazards

considerations determination.

Brief description of amendment: This amendment reflects the

approval of the transfer of the authority to operate the Perry Nuclear

Power Plant, Unit 1, under the license to a new company, FirstEnergy

Nuclear Operating Company. In addition, several administrative changes

unrelated to the transfer are being made to delete certain sections of

the license relating solely to one-time historical events that have

occurred.

Date of issuance: December 21, 1998.

Effective date: December 21, 1998.

Amendment No.: 96.

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

operating license.

Date of initial notice in Federal Register: August 4, 1998 (63 FR

41600).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 21, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear

Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: June 29, 1998, as supplemented

by submittals dated July 14, October 26, and November 30, 1998.

Brief description of amendment: This amendment reflects the

approval of the transfer of the authority to operate Davis-Besse

Nuclear Power Station, Unit 1, under the license to a new company,

FirstEnergy Nuclear Operating Company.

Date of issuance: December 21, 1998.

Effective date: December 21, 1998.

Amendment No.: 228.

Facility Operating License No. NPF-3: Amendment revised the

operating license.

Date of initial notice in Federal Register: August 4, 1998 (63 FR

41602) The additional information provided in the supplemental

submissions provided clarifying information only which did not affect

the staff's proposed no significant hazards consideration or expand the

scope of the application as noticed initially.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 21, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606.

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

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

Date of amendment request: August 2, 1996 (TXX-96434), as

supplemented by

[[Page 2259]]

letters dated October 2, 1998 (TXX-98215), and November 13, 1998 (TXX-

98241 and TXX-98244).

Brief description of amendments: The amendment increases the

allowed outage time (AOT) for a centrifugal charging pump from 72 hours

to 7 days and adds a Configuration Risk Management Program.

Date of issuance: December 29, 1998.

Effective date: December 29, 1998, to be implemented within 30

days.

Amendment Nos.: Unit 1--Amendment No. 62; Unit 2--Amendment No. 48.

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

revised the Technical Specifications.

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

FR 65617) supersedes FR notice dated September 24, 1997.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

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

19497, Arlington, TX 76019.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: May 8, 1998, as supplemented on

July 10 and October 2, 1998.

Brief description of amendment: The amendment reduces the normal

operating suppression pool water temperature limit and adds a time

restriction for the temperature limit allowed during surveillances that

add heat to the suppression pool.

Date of Issuance: December 28, 1998.

Effective date: December 28, 1998, to be implemented within 30

days.

Amendment No.: 163.

Facility Operating License No. DPR-28: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 23, 1998 (63

FR 50941).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated December 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

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

281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: September 12, 1996, as

supplemented April 24, 1997, and September 24, 1998

Brief Description of amendments: The amendments revise License

Condition 3.I, Fire Protection, and relocate fire protection

requirements from the Technical Specifications to the Updated Final

Safety Analysis Report.

Date of issuance: December 16, 1998.

Effective date: December 16, 1998.

Amendment Nos.: 217 and 217.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments

change the Licenses and Technical Specifications.

Date of initial notice in Federal Register: November 4, 1998 (63 FR

59598).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 16, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

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

Project No. 2, Benton County, Washington

Date of application for amendment: October 10, 1996, as

supplemented by letter dated November 9, 1998.

Brief description of amendment: The amendment changes Facility

Operating License No. NPF-21 to authorize the storage of byproduct,

source, and special nuclear materials at the WNP-2 site. These

materials had been originally stored at the WNP-1 site and are not

intended for use at WNP-2.

Date of issuance: December 29, 1998.

Effective date: December 29, 1998, to be implemented within 45 days

from the date of issuance.

Amendment No.: 155.

Facility Operating License No. NPF-21: The amendment revised the

operating license.

Date of initial notice in Federal Register: September 23, 1998 (63

FR 50942).

The November 9, 1998, supplemental letter provided additional

clarifying information that did not change the staff's original no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Dated at Rockville, Maryland, this 6th day of January 1999.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

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

Nuclear Reactor Regulation.

[FR Doc. 99-660 Filed 1-12-99; 8:45 am]

BILLING CODE 7590-01-P