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 March 27 through April 9, 1999. The last

biweekly notice was published on April 7, 1999 (64 FR 17021).

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

[[Page 19555]]

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 May 21, 1999, the licensee may file a request for a hearing with

respect to issuance of the amendment to the subject facility operating

license and any person whose interest may be affected by this

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

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.

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

Michigan

Date of amendment request: March 23, 1999.

Description of amendment request: The proposed amendment would

modify Technical Specification Surveillance Requirement 4.4.1.1.1 to

require each recirculation pump discharge valve to be demonstrated

OPERABLE at least once every 18 months and will delete footnote * that

applies to Technical Specification 4.4.1.1.1.

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 change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes to the Technical Specifications (TS) would

modify the frequency of cycling the recirculation pump discharge

valves from ``each STARTUP*

[[Page 19556]]

prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER'' to

``at least once per 18 months;'' and replace the footnote applicable

to TS 4.4.1.1, ``*If not performed in the previous 31 days'' with

``*Not Used.'' The change in testing frequency does not affect the

probability of an accident since the valve testing is not related to

accident initiation sequences. Consequences of accidents are not

significantly increased because the proposed testing interval

provides reasonable assurance that the valves will function. Testing

of the valves will still be performed on a frequency that is allowed

by TS if no events occur that require entry into Mode 3 or Mode 4.

Therefore, the change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Testing the valves in accordance with the inservice testing (IST)

program on the same testing frequency as testing performed for the

low pressure coolant injection system, provides adequate assurance

that the valves can perform their safety function and will not

increase the consequences of an accident previously evaluated. The

change to the footnote is administrative in nature and will have no

effect on the probability of an accident and will not increase any

safety consequences.

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

different kind of accident from any accident previously evaluated.

The proposed changes revise performing the testing of the

recirculation pump discharge valves from ``prior to Startup* not to

exceed 25% of rated thermal power.'' to ``at least once per 18

months'' and replace the footnote applicable to TS 4.4.1.1'' *If not

performed in the previous 31 days'' with ``*Not Used'' does not

result in a new accident precursor since the test only verifies that

the valve can close which is its safety function. Deleting the

information contained in footnote ``*'' that applies to TS 4.4.1.1.1

and designating it as ``* Not Used.'' is administrative in nature

with no safety significance. Therefore, no different type of

accident from any previously evaluated is introduced.

3. The change does not involve a significant reduction in the

margin of safety.

The proposed changes revise the frequency of cycling the

recirculation pump discharge valves from ``each STARTUP* prior to

THERMAL POWER exceeding 25% of RATED THERMAL POWER'' to ``at least

once per 18 months'' and replace the footnote applicable to TS

4.4.1.1 ``*If not performed in the previous 31 days'' with ``*Not

Used.'' Altering the test frequency does not change valve stroke

time or other performance or design characteristics related to the

safety function of the valves. The potential for failure of the

valve to close is not changed as a result of the proposed change

since the same frequency is allowed by the current TS if no events

occur that require entry into Mode 3 or Mode 4. Performing stroke

time testing on a refueling outage basis and MOV testing on a

periodic basis does not decrease the margin of safety associated

with the valve performing its safety function. Revising footnote *

is an administrative change and has no safety consequence.

Therefore, the proposed changes do 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: Monroe County Library System,

Ellis Reference and Information Center, 3700 South Custer Road, Monroe,

Michigan 48161.

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

2000 Second Avenue, Detroit, Michigan 48226.

NRC Section Chief: George F. Dick, Acting.

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 amendment request: March 3, 1999.

Description of amendment request: The proposed amendments would

change the required qualifications for operations management specified

in the Technical Specifications (TSs) for the Beaver Valley Power

Station, Units 1 and 2 (BVPS-1 and BVPS-2). The requirement that the

operations manager hold a Senior Reactor Operator (SRO) license at the

time of appointment would be changed in the TSs to require that the

assistant operations managers, one for each unit, hold an SRO license

on their assigned unit. The TSs would not then require the operations

manager hold an SRO license. Additionally, the Updated Final Safety

Analysis Report (UFSAR) for each unit would be changed to require the

operations manager to hold, or have held, an SRO license rather than

presently hold a license. The UFSAR would require the same as the TS;

that the assistant operations managers hold an SRO license on the unit

to which they are assigned. Finally, the proposed amendments would

substitute generic personnel titles for plant-specific personnel titles

in the BVPS-1 and BVPS-2 TSs. The correlation between generic titles

and plant-specific titles would be provided in the BVPS-2 UFSAR.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes are administrative in nature. The revised

requirements for who must hold a current senior reactor operator

(SRO) License does not involve any change to the configuration or

method of operation of any plant equipment that is used to mitigate

the consequences of an accident nor alter the conditions or

assumptions in any of the Updated Final Safety Analysis Report

[UFSAR] accident analyses. The requirement that the operations

manager hold or have held an SRO License is included in the revised

Position Qualifications in the Unit 2 UFSAR, Table 13.1-2, sheet 30

of 35. The title changes are being made, consistent with TSTF-65,

Rev 1 and help avoid the need for future Technical Specification

changes. Therefore, it can be concluded that the proposed changes do

not involve any increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

No new failure modes are defined for any plant system or

component important to safety nor has any new limiting failure been

identified as a result of the proposed changes. Therefore, it can be

concluded that the proposed change does not create the possibility

of a new or different kind of accident from those previously

evaluated.

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

of safety?

The proposed changes are administrative in nature. One of the

proposed changes requires that the manager who directly supervises

the licensed operators at each unit be the holder of a current SRO

license. The other change modifies personnel titles. Therefore, it

can be concluded that the proposed changes do not involve any

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: B.F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Section Chief: Singh Bajwa.

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

Station, Unit No. 2, Shippingport, Pennsylvania

Date of amendment request: March 16, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.7.1.3

[[Page 19557]]

and associated Bases for the Primary Plant Demineralized Water (PPDW)

System to clarify that the minimum specified volume of water in the

PPDW Storage Tank is a usable volume. Additionally, the minimum usable

volume of water in the PPDW Storage tank is increased, and a clarifying

footnote that the specified value is an analysis value is added.

Finally, several editorial and administrative changes, such as revision

of action statement wording, addition of license number to the TS page,

and addition of clarifying information to the TS Bases regarding

analysis assumptions are made.

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 failure of the primary plant demineralized water (PPDW)

storage tank to provide a sufficient source of water to the

Auxiliary Feedwater (AFW) System is not an accident initiating

event. Therefore, the probability of an accident previously

evaluated is not increased by this proposed amendment.

Limiting Condition for Operation (LCO) 3.7.1.3 titled ``Primary

Plant Demineralized Water (PPDW)'' will be revised to specify the

required value for PPDW storage tank volume as a usable volume. To

reflect the value currently assumed in the analysis, the value

stated in the LCO, for minimum required PPDW storage tank volume,

would be slightly increased. The addition of proposed Footnote (1)

to LCO 3.7.1.3 will ensure that plant operators recognize that the

specified volume is an analysis value and that the value does not

include measurement uncertainties. This footnote will require plant

procedures to specify an increased required volume in the PPDW

storage tank to account for measurement uncertainties. The proposed

revisions to LCO 3.7.1.3 will assure that the PPDW storage tank

minimum usable volume is maintained consistent with the design basis

for the PPDW storage tank. The PPDW storage tank will continue to

provide a sufficient source of water to the AFW pumps. Maintaining a

sufficient source of water will ensure that the AFW System is

capable of mitigating the consequences of Design Basis Accidents

(DBAs) that could result in overpressurization of the RCS pressure

boundary. The AFW system will continue to be capable of providing an

emergency source of feedwater to the steam generators to act as heat

sinks for sensible and decay heat removal from the reactor core. A

sufficient volume of water will continue to be maintained in the

PPDW storage tank to satisfy the Safe Shutdown evaluation.

The proposed changes to the Action statements will remove the

required water volume value and add wording pertaining to the water

volume not being within the limit. The LCO clearly states the value

for the minimum required volume in the PPDW storage tank. Therefore,

the proposed modification to the Action statements is administrative

in nature and does not affect plant safety. The additional Bases

wording pertaining to reactor coolant pump operation is

administrative in nature and does not affect plant safety. The

remaining change, which consists of the addition of plant operating

license number, is editorial in nature and does not affect plant

safety.

Therefore, operation of the facility in accordance with the

proposed amendment does not involve a significant increase in the

probability or consequence of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed amendment will not change the physical plant or the

modes of plant operation defined in the operating license. This

change does not involve the addition or modification of plant

equipment nor does it alter the design or operation of plant

systems. The proposed amendment will require that the minimum volume

in the PPDW storage tank be maintained consistent with analysis

assumptions.

Therefore, operation of the facility in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

of safety?

The minimum required volume in the PPDW storage tank would be

slightly increased over the currently required value. This increase

in the required volume will ensure that an adequate volume of water

is maintained in the PPDW storage tank. The proposed addition of the

term ``usable,'' along with the addition of Footnote (1), will

ensure that the water volume specified in LCO 3.7.1.3 is

appropriately increased in plant procedures to account for unusable

volume in the tank and for measurement uncertainties. A sufficient

volume of water will continue to be maintained in the PPDW storage

tank to satisfy the Safe Shutdown evaluation.

The PPDW storage tank will continue to provide a sufficient

source of water to the AFW pumps to ensure that the AFW System is

capable of mitigating the consequences of DBAs that could result in

overpressurization of the RCS pressure boundary. The AFW system will

continue to be capable of providing an emergency source of feedwater

to the steam generators to act as heat sinks for sensible and decay

heat removal from the reactor core.

The proposed changes to the Action statements will remove the

required water volume value and add wording pertaining to the water

volume not being within the limit. The LCO clearly states the value

for the minimum required volume in the PPDW storage tank. Therefore,

the proposed modification to the Action statements is administrative

in nature and does not affect plant safety. The additional Bases

wording pertaining to reactor coolant pump operation is

administrative in nature and does not affect plant safety. The

remaining change, which consists of the addition of plant operating

license number, is editorial in nature and does not affect plant

safety.

Therefore, the proposed amendment does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Section Chief: S. Singh Bajwa.

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

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

Date of amendment request: August 31, 1998, and revised March 18,

1999.

Description of amendment request: The proposed amendment would

revise Improved Technical Specification (ITS) 5.6.2.10, ``Steam

Generator (OTSG [once-through steam generator]) Tube Surveillance

Program,'' to include a new repair process, called a ``repair roll'' or

``re-roll.'' The process would be used to repair steam generator tubes

with defects within the upper tubesheet. Changes to inservice

inspection and reporting requirements are proposed for tubes which are

repaired using this process. In addition, several format and editorial

changes are proposed to ITS 5.6.2.10 and to ITS 5.7.2, ``Special

Reports,'' for clarification purposes. The March 18, 1999 revision

superceded the August 31, 1998 request, and includes the results of

recent accident analyses conducted to identify the maximum OTSG tube

tensile loads. As a result of the increased tube tensile loads, some

tubes will require a double repair roll. The double repair roll

methodology was not included in the original amendment request.

Therefore, this notice revises the previous Notice of Consideration of

Issuance of Amendment (63 FR 56249).

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.

[[Page 19558]]

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated?

The repair roll process is a method to create a new primary-to-

secondary pressure boundary joint in the upper tubesheet of Babcock

& Wilcox (B&W) Once Through Steam Generators (OTSGs) manufactured

with Inconel Alloy 600 tubes. The repair roll process creates a new

roll joint in the OTSG tubes at a point closer to the secondary face

of the tubesheet than the existing roll joint. The new pressure

boundary is established by the repair roll to remove degradation of

the existing roll joint from pressure boundary service. The repair

roll process has been qualified as an acceptable repair methodology

for use in the upper tubesheet of the Crystal River Unit 3 (CR-3)

OTSGs. The proposed License Amendment Request (LAR) proposes to

implement the qualified OTSG tube repair roll process, and also

addresses several editorial and format changes which do not impact

the current CR-3 accident analyses.

The qualification of the OTSG tube repair roll methodology is

based on establishing a mechanical joint length that will carry all

structural loads imposed on the OTSG tubes while maintaining the

required margins during normal and accident conditions. A series of

tests and analyses were performed to establish the minimum

acceptable length of the OTSG tube repair roll. Tests performed

included leak, tensile, fatigue, ultimate load and eddy-current

measurement uncertainty. The analyses evaluated plant operating and

faulted load conditions, in addition to OTSG tubesheet bow effects.

OTSG tube leakage remains bounded by the evaluation presented in the

CR-3 Final Safety Analysis Report (FSAR) for a main steam line break

(MSLB). The proposed change also includes a description of the

required inspection program for the OTSG tube repair rolls. The

additional inspection requirements do not change any accident

initiators. The proposed inspections following OTSG tube repair roll

installation, and during future inservice inspections, assure

continuous monitoring of these tubes such that inservice degradation

of tubes repaired by the repair roll process will be detected. Based

on the qualification testing and analyses performed, as well as the

industry experience with the use of OTSG tube repair roll processes,

there are no new safety issues associated with the use of repair

roll methodology. Therefore, this change does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

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

accident from previously evaluated accidents?

The repair roll creates no new failure modes or accident

scenarios. The new pressure boundary joint created by the repair

roll process has been demonstrated, by testing and analysis, to

provide structural and leakage integrity equivalent to the original

design and construction for all normal operating and accident

conditions. Furthermore, the testing and analysis demonstrate the

repair roll process creates no new adverse effects for the repaired

tube and does not change the design or operating characteristics of

the OTSGs. In the unlikely event that a tube with a repair roll

should fail and sever completely at the transition of the repair

roll region, the tube would remain engaged in the tubesheet bore,

preventing interaction with other surrounding tubes. In this case,

leakage is bounded by the steam generator tube rupture (SGTR)

accident analysis. Therefore, this change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

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

The repair roll process effectively removes the defective/

degraded area of the tube from service. The repair roll interface

created with the tubesheet satisfies the necessary structural,

leakage and heat transfer requirements. The mechanical joint is

constrained within the tubesheet bore; thus, there is no additional

risk associated with tube rupture. The accident leakage is shown to

be less than one gallon per minute primary-to-secondary leakage.

Therefore, the FSAR analyzed accident scenarios remain bounding, and

the use of the repair roll process does not reduce the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Section Chief: Sheri R. Peterson.

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

1, DeWitt County, Illinois

Date of amendment request: March 1, 1999.

Description of amendment request: The proposed amendment would

approve changes to the Updated Safety Analysis Report (USAR) concerning

design requirements for physical protection from tornado missiles for

safety-related equipment.

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 activity does not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

The associated USAR changes reflect use of the Electric Power

Research Institute (EPRI) Topical Report, ``Tornado Missile Risk

Evaluation Methodology, (EPRI NP-2005),'' Volumes I and II. This

methodology has been reviewed, accepted and documented in an NRC

Safety Evaluation dated October 26, 1983. The NRC concluded that:

``the EPRI methodology can be utilized when assessing the need for

positive tornado missile protection for specific safety-related

plant features in accordance with the criteria of SRP Section

3.5.1.4.''

The EPRI methodology has been previously applied at CPS to

resolve previously identified missile protection issues during the

initial licensing of the plant. The NRC documented their acceptance

of this methodology in Supplement 6 to the CPS Safety Evaluation

Report (NUREG-0853, July 1986).

As permitted in the Standard Review Plan (NUREG-0800), the total

probability of damage to plant systems or components initiated from

tornado missiles leading to consequences in excess of 10 CFR Part

100 guidelines will be maintained below an acceptable level. The

results of the current tornado missile hazards analysis are such

that the calculated total tornado missile hazard probability is

approximately 3.4 x 10-7 per year. This is lower than the value

determined to be acceptable, i.e., 1 x 10-6 per year.

Although it has been calculated that these targets have a higher

total probability of being exposed to tornado missiles than that

described to be acceptable in SER Supplement 6, Section 3.5.1.3, the

revised tornado missile hazards analysis for CPS has determined that

this probability is acceptably low.

With respect to the probability of occurrence or the

consequences of an accident previously analyzed in the USAR, the

possibility of a tornado reaching CPS and causing damage to plant

systems, structures and components is a design basis event

considered in the USAR. The changes being proposed herein do not

affect the probability that a tornado will reach the plant, but they

do, from a licensing basis perspective, reflect a slightly

increased, calculated probability that missiles generated by the

winds of a tornado might strike certain plant systems or components.

The tornado missile analysis determined that there are a limited

number of safety-related components that theoretically could be

struck. The probability of tornado-generated missile strikes on

important systems and components (as discussed in Regulatory Guide

1.117) was analyzed using the probability methods described above.

Based on the low, calculated probability, the total (cumulative)

probability of strikes will be maintained below an adequately low

acceptance criterion to ensure overall plant safety. On this basis,

the proposed change is not considered to constitute a significant

increase in the probability of occurrence or the consequences of an

accident, due to the low probability of a tornado missile striking

safety-related systems or components.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of previously evaluated

accidents.

[[Page 19559]]

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes involve evaluation of whether any physical

protection of safety-related equipment from tornado missiles is

required relative to the probability of such damage without physical

protection. A tornado at CPS is a design basis event considered in

the USAR, however, a tornado is not postulated to act as an

initiator for any new or different kind of accident, or to occur

coincident with any of the design basis accidents in the USAR. The

low probability threshold established for missile damage to plant

systems is consistent with these assumptions.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident.

(3) The proposed activity does not involve a significant

reduction in a margin of safety.

Under the proposed change, physical protection of safety-related

equipment from tornado missiles must be considered if it has been

determined that the calculated total tornado missile hazard

probability is greater than 1 x 10-6 per year. The proposed change

to the USAR to specifically identify this threshold may slightly

increase the probability of a malfunction of equipment important to

safety previously evaluated in the safety analysis report (i.e.,

changing the requirements from protecting all safety-related systems

and components to not requiring protection if there is an extremely

low probability that a tornado missile could strike portions of

safety related systems and components). However, the changes are

consistent with the minimum acceptable requirements as documented in

the NRC's Safety Evaluation Report dated October 23, 1983.

Therefore, there will be no significant reduction to the margin of

safety that may be associated with the potential for safety-related

equipment to be damaged from tornado-generated missiles.

Therefore, the proposed changes do 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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, IL 61727.

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th Street,

Decatur, IL 62525.

NRC Section Chief: Anthony J. Mendiola.

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: March 17, 1999.

Description of amendment request: The licensee is proposing to

change Technical Specifications 3.5.2, ``Emergency Core Cooling

Systems--ECCS Subsystems--Tavg greater than or equal to 300 deg.F;''

3.7.1.7, ``Plant Systems--Atmospheric Steam Dump Valves;'' and 3.7.6.1,

``Plant Systems--Control Room Emergency Ventilation System.'' The

proposed Technical Specification changes will revise (1) surveillance

requirements for Emergency Core Cooling System valves, (2) the

atmospheric steam dump valve requirements to focus on the steam release

path instead of the individual valves, and (3) the allowed outage times

for the atmospheric steam dump valves and Control Room Emergency

Ventilation System.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with 10 CFR 50.92, NNECO 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.

Technical Specification 3.5.2

The removal of 2-CH-434, a manual valve, from the list of valves

to be checked every 31 days by Surveillance Requirement (SR)

4.5.2.a.10 will not change the requirement for this containment

isolation valve to be locked closed. The position of valve 2-CH-434,

and the associated locking device, will be verified by SR 4.6.1.1.a.

Although this change will result in the position of 2-CH-434 being

checked less often, there are sufficient Technical Specification and

administrative requirements to ensure that 2-CH-434 will be

maintained in the proper position. An additional benefit of this

proposed change will be a reduction in personnel exposure since 2-

CH-434 is located inside containment. This proposed change will not

result in any modification to Emergency Core Cooling System (ECCS)

alignment or operation.

The addition of the footnote to SR 4.5.2.a.10 will clarify that

2-SI-306 is pinned and locked open to the required throttle

position. 2-SI-306, which is the Shutdown Cooling (SDC) System

throttle valve in the discharge piping of the SDC pumps, is required

to be left in a throttled position after SDC has been secured to

ensure sufficient low pressure safety injection (LPSI) flow will be

available. This proposed change will not result in any modification

to ECCS alignment or operation.

The change in the valve nomenclature used in SR 4.5.2.e and

Table 4.5-1 from throttle valve to injection valve will eliminate

any confusion between valve description and valve operation. This

proposed change will not result in any modification to ECCS

alignment or operation.

The addition of the License Amendment Number to the bottom of

Page 3/4 5-6a will not result in a technical change to this

Technical Specification.

Technical Specification 3.7.1.7

The proposed changes will expand the scope of Technical

Specification 3.7.1.7 to include the steam release path, instead of

just the individual atmospheric dump valves (ADVs). The allowed

outage times will be modified to address inoperable ADV lines and

the impact inoperable ADV lines will have on the ability of

Millstone Unit No. 2 to mitigate a loss of coolant accident (LOCA).

If one ADV line is inoperable, a plant shutdown will be required if

the ADV line is not restored to operable status within 48 hours. An

allowed outage time of 48 hours to restore the ADV line to operable

status is acceptable based on the low probability of a LOCA

occurring during this time period, and the subsequent loss of

offsite power and the failure of one train of high pressure safety

injection (HPSI). This is also consistent with the allowed outage

time for one ECCS train (Technical Specification 3.5.2).

If two ADV lines are inoperable, a plant shutdown will be

required if at least one ADV line is not restored to operable status

within one hour. The plant will be required to be in Mode 3 within

the following 6 hours. These time requirements are based on

Technical Specification 3.0.3. However, the time to reach Mode 4

will remain at the ``following 24 hours'' to reflect the impact

inoperable ADV lines may have on the time to cool down the plant.

The proposed change to the surveillance requirement will ensure

operation of the ADV lines, consistent with the accident analysis,

is verified.

The proposed change in component nomenclature is consistent with

current Millstone Unit No. 2 terminology. This is not a technical

change.

The proposed changes to the Bases of Technical Specification

3.7.1.7 are consistent with the changes just described.

Technical Specification 3.7.6.1

The action requirements for the Control Room Emergency

Ventilation System will be modified to address the situation when

both Control Room Emergency Ventilation Trains are inoperable in

Modes 1, 2, 3, and 4. This situation is expected to occur during

normal plant operation when the air filters in the common supply

header to both trains are cleaned/replaced. Since this is a common

supply header, both trains are affected and would be inoperable. The

proposed action requirements will address this situation so

[[Page 19560]]

that Technical Specification 3.0.3 will not be entered as a result

of an expected plant activity. However, since the proposed action

requirements are the same as the requirements of Technical

Specification 3.0.3, the time the plant is allowed to operate in

this situation will not change.

The proposed changes to the Technical Specifications and

associated Bases will have no adverse effect on plant operation or

accident mitigation equipment. The proposed changes will ensure that

the necessary equipment to mitigate the design basis accidents will

be available, or a plant shutdown will be required. In addition, the

proposed changes can not cause an accident, and they will ensure the

accident mitigation equipment will continue to operate as assumed in

the analyses to mitigate the design basis accidents. Therefore,

there will be no 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 changes to the Technical Specifications and

associated Bases will have no adverse effect on plant operation or

accident mitigation equipment. The proposed changes will ensure that

the necessary equipment to mitigate the design basis accidents will

be available, or a plant shutdown will be required. Therefore, the

proposed changes 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 changes to the Technical Specifications and

associated Bases will ensure that the necessary equipment to

mitigate the design basis accidents will be available, or a plant

shutdown will be required. The proposed changes will not result in

any plant configuration changes. There will be no adverse effect on

plant operation or accident mitigation equipment. The plant response

to the design basis accidents will not change. Therefore, there will

be no significant reduction in the margin of safety as defined in

the Bases for the Technical Specifications affected by these

proposed changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Section Chief: James W. Clifford.

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: March 19, 1999.

Description of amendment request: The proposed changes will

relocate Technical Specifications (TSs) 3.3.3.2, ``Instrumentation,

Incore Detectors,'' 3.3.3.3, ``Instrumentation, Meteorological

Instrumentation,'' to the Millstone, Unit No. 2 Technical Review Manual

(TRM). Index Page V will be revised by eliminating the sections

corresponding to incore detectors (Page \3/4\ 3-0), seismic

instrumentation (Page \3/4\ 3-32), and meteorological instrumentation

(Page \3/4\ 3-36). These sections, as well as changes to the associated

Bases, will be relocated to the TRM.

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 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.

Technical Specification 3.3.3.2, Instrumentation, ``Incore

Detectors,'' is proposed to be relocated to the TRM where future

changes will be controlled in accordance with 10 CFR 50.59.

Relocation of this Technical Specification to the TRM does not imply

any reduction in its importance in confirming that core power

distribution are bounded by safety analysis limits. These

instruments are neither used for, nor capable of, detecting a

significant abnormal degradation of the reactor coolant pressure

boundary before a design basis accident, nor do they function as a

primary success path to mitigate events which assume a failure of,

or a challenge to, the integrity of fission product barriers.

Although the core power distribution (measured by the incore

detectors) constitutes an important initial condition to design

basis accidents and therefore needs to be addressed by Technical

Specifications, the detectors themselves are not an active design

feature needed to preclude analyzed accidents or transients. The

proposed change will not alter the way core power distribution is

measured by the incore detectors, nor will it alter any of the power

distribution assumptions used in the accident analysis. Therefore,

this change will not significantly increase the probability or

consequences of an accident previously evaluated.

Technical Specification 3.3.3.3, Instrumentation, ``Seismic

Instrumentation,'' is proposed to be relocated to the TRM where

future changes will be controlled in accordance with 10 CFR 50.59.

Relocation of Technical Specification 3.3.3.3 to the TRM does not

imply any reduction in its importance in determining the response of

those nuclear power plant features important to safety in the event

of an earthquake. Seismic instrumentation does not actuate any

protective equipment or serve any direct role in the mitigation of

an accident. The capability of the plant to withstand a seismic

event or other design basis accident is determined by the initial

design and construction of systems, structures, and components. The

instrumentation is used to alert operators to the seismic event and

evaluate the plant response. The seismic instrumentation does not

serve as a protective design feature or part of a primary success

path for events which challenge fission product barriers. The

proposed change will not alter the way these instruments are used in

determining the response of those nuclear power plant features

important to safety in the event of an earthquake, nor will it alter

the capability of the plant to withstand a seismic event. Therefore,

this change will not significantly increase the probability or

consequences of an accident previously evaluated.

Technical Specification 3.3.3.4, Instrumentation,

``Meteorological Instrumentation,'' is proposed to be relocated to

the TRM where future changes will be controlled in accordance with

10 CFR 50.59. Relocation of Technical Specification 3.3.3.4 to the

TRM does not imply any reduction in its importance in providing a

basis for estimating annual radiation doses resulting from

radioactive materials released in airborne effluents. The

instrumentation does not serve to ensure that the plant is operated

within the bounds of initial conditions assumed in design basis

accident and transient analyses or that the plant will be operated

to preclude transients or accidents. Likewise, the meteorological

instrumentation does not serve as part of the primary success path

of a safety sequence analysis used to demonstrate that the

consequences of these events are within the appropriate acceptance

criteria. The proposed change will not alter the way these

instruments are used in providing a basis for estimating annual

radiation doses resulting from radioactive materials released in

airborne effluents. Therefore, this change will not significantly

increase the probability or consequences of an accident previously

evaluated.

Revision of Index page V and the proposed changes to the

associated Bases sections are administrative changes. Therefore,

these changes will not significantly increase the probability or

consequences of an accident previously evaluated.

The proposed changes do not alter how any structure, system, or

component functions. There will be no effect on

[[Page 19561]]

equipment important to safety. The proposed changes have no effect

on any of the design basis accidents previously evaluated.

Therefore, this License Amendment Request does not impact the

probability of an accident previously evaluated, nor does it involve

a significant increase in the 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 changes do not alter the plant configuration (no

new or different type of equipment will be installed) or require any

new or unusual operator actions. They do not alter the way any

structure, system, or component functions and do not alter the

manner in which the plant is operated. The proposed changes do not

introduce any new failure modes. Therefore, the proposed changes

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 relocation of incore detector instrumentation

requirements to the TRM does not imply any reduction in their

importance in confirming that core power distribution is bounded by

safety analysis limits. The incore detectors will still be used to

measure core power distribution and the assumptions used in the

accident analysis will be verified. The proposed relocation of

seismic instrumentation requirements to the TRM does not imply any

reduction in their importance in determining the response of those

nuclear power plant features important to safety in the event of an

earthquake. The seismic instrumentation will still be used to

determine the response of those nuclear power plant features

important to safety in the event of an earthquake. The capability of

the plant to withstand a seismic or other design basis accident,

which is determined by the initial design and construction of

systems, structures, and components will not be altered. The

relocation of meteorological instrumentation requirements to the TRM

does not change the way these instruments are used in providing a

basis for estimating annual radiation doses resulting from

radioactive materials released in airborne effluents. The

meteorological instrumentation will continue to perform their

function in exactly the same way.

The proposed changes do not affect any of the assumptions used

in the accident analysis, nor do they affect any operability

requirements for equipment important to plant safety. Therefore, the

proposed changes will not result in a significant reduction in the

margin of safety as defined in the Bases for Technical

Specifications covered in this License Amendment Request.

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 Section Chief: James W. Clifford.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of amendment request: September 11, 1998, as supplemented by

letter dated January 14, 1999.

Description of amendment request: The proposed amendments would

change the combined Technical Specifications (TS) for the Diablo Canyon

Power Plant, Unit Nos. 1 and 2 to revise TS 6.8.4f., ``Containment

Polar and Turbine Building Cranes,'' to control the operation of the

containment polar cranes in jet impingement zones.

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 Technical Specification (TS) 6.8.4f requirement to have a

program that will ensure the position of the polar cranes precludes

jet impingement from a postulated pipe rupture was previously

evaluated in the NRC staff's safety evaluation for License

Amendments (LA) 20 and 21. The proposed change is to control the

operation of the containment polar cranes in jet impingement zones.

PG&E evaluated a high energy line break (HELB) scenario for core

damage frequency (CDF) considering operation of a polar crane. A

postulated HELB would have to damage the crane or cause its load to

drop in a manner that damages a component that exacerbates the HELB

event and leads to core damage. The PRA evaluation for this scenario

concluded the CDF is 1.6E-9 per year. It is not a significant

increase in CDF compared to never operating the polar crane in jet

impingement zones. The CDF for this scenario is nonrisk significant

when compared to the industry standard threshold for risk

significance for an operational evolution, which is 1E-6 per year.

Several factors that further lower the risk of CDF include: 1) the

movement of heavy loads is done in accordance with the DCPP Heavy

Loads Program, which provides assurance that a dropped load would

not lead to core damage, 2) the polar crane had been evaluated to

withstand jet impingement loads without the seismic loads, and 3)

the probability of simultaneous seismic and HELB events is low.

Therefore, based on probabilistic considerations, the risk

associated with this proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

Deterministic engineering methods required combining both the

seismic and jet impingement loads to qualify Design Class I

structures. The polar cranes were not originally qualified for these

combined loads. This resulted in administrative controls that

prohibited parking the polar cranes in jet impingement zones to

preclude jet impingement loads from a postulated pipe rupture. The

proposed change does not involve a physical change to the plant, but

it does involve a change to the TS required program for containment

polar crane operation.

The proposed change is to control the operation of the

containment polar cranes in jet impingement zones. It recognizes

that there are jet (HELB) and target (polar crane) interactions.

They were previously not considered for postulated jet impingement

analyses because administrative controls prohibited parking the

polar cranes in jet impingement zones. PG&E has evaluated jet

impingement loads on the polar crane and determined it is able to

withstand these loads without seismic loads. Based on this

evaluation, the polar crane would not fail due to a HELB event. The

movement of a heavy load would be done in accordance with the DCPP

Heavy Loads Program. Thus, there would be no consequential failures

that would lead to core damage.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The current TS 6.8.4f. requirement to have a program that will

ensure the position of the polar cranes precludes jet impingement

from a postulated pipe rupture was previously evaluated in the NRC

staff's safety evaluation for LAs 20 and 21.

The credible HELB sources that could impinge on the polar crane

were identified and evaluated. The feedwater and main steam line

steam generator nozzles are the only credible HELBs that could

impinge upon the polar crane. The structural integrity of these

lines was evaluated and determined to be of robust design.

The margin of safety affected by the proposed change involves a

comparison between the margin of safety afforded by no operation of

the polar crane and operation that is controlled by procedures. The

margin of safety in this case is the increase in risk for CDF caused

by a scenario that postulates that operation of the polar crane

would lead

[[Page 19562]]

to core damage. The risk for CDF has been evaluated and determined

to be nonrisk significant. The CDF value is well below the industry

standard threshold for acceptable risk for an operational evolution,

which is 1E-6 per year.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room Location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407.

Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas &

Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Project Director: Stuart A. Richards.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of amendment request: December 12, 1998.

Description of amendment request: The proposed amendments would

change the combined Technical Specifications (TS) for the Diablo Canyon

Power Plant, Unit Nos. 1 and 2 to revise TS 6.9.1.8, ``Core Operating

Limits Report,'' to allow use of NRC approved addenda to WCAP-10054-P-

A, ``Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP

Code,'' August 1985, to determine core operating limits.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This change is administrative in nature in that it revises the

Technical Specification (TS) Administrative Controls for the Core

Operating Limits Report to include reference to NRC approved addenda

to WCAP-10054-P-A, ``Westinghouse Small Break ECCS Evaluation Model

Using the NOTRUMP Code,'' August 1985. The proposed change would

allow the use of the analytical methods in WCAP-10054-P-A, Addendum

2, Revision 1, Addendum to the Westinghouse Small Break ECCS.

Evaluation Model Using the NOTRUMP Code: Safety Injection Into

the Broken Loop and COSI Condensation Model,'' July 1997, and other

NRC approved addenda to WCAP-10054-P-A to determine core operating

limits for Diablo Canyon Power Plant (DCPP). Because plant operation

will continue to be limited in accordance with cycle specific core

operating limits that are established using an NRC approved

methodology, NRC approved addenda to WCAP-10054-P-A are acceptable

for use in determining DCPP Unit 1 and 2 cycle specific core

operating limits.

The change does not affect plant operation, or physically alter

or change the function of structures, systems, or components

required to mitigate the consequences of a design basis accident. In

addition, it cannot initiate a transient or affect the probability

of occurrence of any previously analyzed accident.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change revises the TS to allow the use of NRC

approved analytical methods in WCAP-10054-P-A, Addendum 2, Revision

1, and other NRC approved addenda to WCAP-10054-P-A, to determine

core operation limits. The change is consistent with the

requirements of the TS, and does not affect plant operation, or

physically alter or change the function of structures, systems, or

components required to mitigate the consequences of a design basis

accident.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change revises the TS to allow the use of the NRC

approved analytical methods in WCAP-10054-P-A, Addendum 2, Revision

1 and other NRC approved addenda to WCAP-10054-P-A, to determine

core operating limits. The change is consistent with the

requirements of the TS, and does not affect plant operation, or

physically alter or change the function of structures, systems, or

components required to mitigate the consequences of a design basis

accident. The acceptance limits for the small break loss-of-coolant

accident are not affected by this change and will continue to be

met.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room Location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407.

Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas &

Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Project Director: Stuart A. Richards.

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 amendment request: January 25, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to

relocate the time restriction for movement of irradiated fuel and its

related basis page from the TSs to the IP3 Final Safety Analysis Report

(FSAR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously [evaluated]?

Response

Relocation (i.e., removal from TS) of TS 3.8.A.9 and its basis

for the minimum time prior to movement of more than 76 irradiated

fuel assemblies (267 hour limit) will not involve a significant

increase in the probability or consequences of an accident since the

relocation of the TS to administrative controls governed by 10 CFR

50.59 (FSAR) does not affect the availability or function of fuel

storage and handling equipment or the SFP [spent fuel pool] cooling

system. The waiting time of 267 hours following plant shutdown

before unloading more than 76 assemblies from the reactor is to

ensure that the maximum SFP water temperature will be within design

objectives as stated in the FSAR.

The waiting time of 267 hours is not an initiator of an accident

and the proposed change does not alter overall system operation,

physical design, system configuration, or operational setpoints.

There will be no significant increase in the consequences of an

accident because the restricted movement time for irradiated fuel

will continue to be administratively controlled under 10 CFR 50.59.

The other TS of section 3.8.A (such as the remaining portion of

3.8.A.9, and 3.8.A. 10) and the other controls ensure that doses

from a postulated FHA are within 10 CFR 100 limits.

[[Page 19563]]

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response

The basis for the waiting time of 267 hours following plant

shutdown before unloading more than 76 assemblies from the reactor

is to ensure that the maximum pool water temperature will be within

design objectives as stated in the FSAR. Relocation of this waiting

time of 267 hours for irradiated fuel will not create the

possibility of a new or different kind of accident from any

previously evaluated. The TS change will not create the possibility

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

since it does not alter the administrative controls for fuel

handling or the operation, physical design, system configuration, or

operational setpoints for fuel handling and SFP cooling. The plant

systems for fuel storage and handling, and SFP cooling are operated

in the same manner as before and, consequently, the relocation does

not introduce any new accident initiators or failure mechanisms and

does not invalidate the existing FHA response. The minimum waiting

time for movement of more than 76 irradiated fuel assemblies is not

an accident initiator. The minimum waiting time will continue to be

controlled under 10 CFR 50.59.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response

Relocation (i.e., removal from TS) of TS 3.8.A.9 and its basis

for the waiting time of 267 hours following plant shutdown for

irradiated fuel will not involve a significant reduction in margin

of safety. The waiting time of 267 hours following plant shutdown

before unloading more than 76 assemblies from the reactor is to

ensure that the maximum SFP water temperature will be within design

objectives as stated in the FSAR. The relocation is a change to the

administrative controls that are used to limit the heat load on the

SFP cooling system, and those administrative controls will be

governed by 10 CFR 50.59. The manner in which fuel storage and

handling is performed, and how the SFP cooling system is operated

does not change and there is no change to physical design, system

configuration, or operational setpoints. The other controls and the

existing TS assure that dose from a postulated FHA are within 10 CFR

100 limits. Previous analyses remain unchanged. The current TS does

not meet the criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in the

Technical Specifications.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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 amendment request: January 28, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to

change the setpoint of the automatic reactor trip on turbine trip to at

or below the P-8 setpoint.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

The addition of reactor trip on turbine trip at [greater than or

equal to] 50% to the P-8 Permissive function versus its current

setting of [greater than or equal to] 10%, as revised in TS section

2.3.1.C.(3), 2.3.2.A, 2.3.2.B, Table 3.5-2, item 12, Table 4.1-1,

item 21 and associated bases, does not significantly increase the

probability or consequences of an accident previously evaluated.

This additional function, change in reactor trip on turbine trip

setpoint, does not cause the initiation of any accident, nor create

any new credible limiting single failure, nor result in any event

previously deemed incredible being made credible. The existing

separation of the reactor and protection functions are not adversely

impacted. In addition, the safety functions of safety related

systems and component, which are related to accident mitigation,

have not been altered. The change in the P-7 or P-8 circuitry does

not directly initiate an accident. The consequences of accidents

previously [evaluated] in the IP3 FSAR [final safety analysis

report] are unaffected by this change because no change to any

equipment response or accident mitigation scenario has resulted.

There are no additional challenges to fission product barrier

integrity. Therefore, the probability or consequences of an accident

previously evaluated will not be increased.

(2) Does the proposed license amendment create the possibility

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

evaluated?

By adding the reactor trip on turbine trip at [greater than or

equal to] 50% to the P-8 Permissive function and setpoint, versus

its current setting of [greater than or equal to] 10% and revising

TS sections 2.3. l.C.(3), 2.3.2.A, 2.3.2.B, Table 3.5-2, item 12,

Table 4.1-1, item 21 and associated bases, does not create the

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

already evaluated. The additional function added to the P-8

Permissive does not result in any event previously deemed incredible

being made credible. No new accident scenarios, failure mechanisms,

or limiting single failures are introduced as a result of this

change. In addition, the safety functions of safety related systems

and components, which are related to accident mitigation, have not

been altered. Therefore, the possibility of a new or different kind

of accident is not created.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

The addition of the reactor trip on turbine trip at [greater

than or equal to] 50% to the P-8 Permissive function, versus its

current setting of [greater than or equal to] 10% and associated

changes to TS Sections 2.3. l.C.(3), 2.3.2.A, 2.3.2.B, Table 3.5-2,

item 12, Table 4.1-1, item 21 and the associated bases, will have no

effect on the availability, operability or performance of the

safety-related systems and components and does not affect the plant

TS requirements. The current licensing basis safety analyses for IP3

remain bounding with the modification to the P-8 Permissive

function; therefore, the margin of safety as defined in the TS is

not reduced. The change to the IP3 TS does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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 amendment request: January 28, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 Technical Specifications (TSs) proposes to reduce

the number of Emergency Diesel Generators (EDGs) required to be

operable during cold shutdown from 2 to 1 under certain conditions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the

[[Page 19564]]

probability or consequences of an accident previously [evaluated]?

Response

No. The equipment, which is affected by the proposed Technical

Specification change, is not an initiator to those accidents

postulated to occur during Cold Shutdown or Refueling operating

conditions. A comprehensive systems review and EDG loading

electrical analysis has demonstrated the ability of those shutdown

support systems, necessary to provide safe shutdown needs, to

perform their accident mitigation functions for the postulated

accidents during Cold Shutdown and Refueling conditions. One EDG can

support the necessary electrical loads required in Cold Shutdown and

Refueling in the event of postulated accidents along with a LOOP

[loss of offsite power] in the time frame required to prevent

reactor core/cavity/SFP [spent fuel pool] heatup concerns. This EDG

support relies upon existing plant designed manual closure of 480VAC

EDS [electrical distribution system] bus tie breakers to allow a

single EDG to pick up other 480VAC EDS bus loads, such as supplying

an RHR [residual heat removal] pump and SFP cooling pump, located on

480VAC EDS buses 3A, 5A, or 6A. Together, operability of the

required offsite circuit(s) and one EDG ensures the availability of

sufficient AC sources to operate the unit in a safe manner and to

mitigate the consequences of postulated accidents during shutdown

(e.g., Fuel Handling Accidents). Action statements provide prompt,

specific guidance to ensure sufficiently conservative plant response

should the expected EDG power supply not be available. These Action

Statements are similar to those in the STS [Standard Technical

Specifications]. Therefore, the proposed license amendment (i.e.,

changes to 3.7.F.4 and the added sections of 3.7.F.5 & 3.7.F.6) does

not involve a significant increase in the probability or

consequences of an accident previously analyzed.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response

No. The proposed license amendment does not involve any physical

changes to plant systems or component set points. The use of 480VAC

EDS bus tie breakers to power loads from an energized 480VAC bus is

part of present plant design and included within the present LOOP

Off-Normal operating procedures when the reactor is in Cold Shutdown

operating conditions. As discussed in the Standard Technical

Specifications, NUREG 1431, during plant shutdown with one EDG, it

is not required to assume a single failure and concurrent loss of

all offsite or all onsite power. Worst case bounding events are

deemed not credible in Cold Shutdown and Refueling conditions

because the energy contained within the reactor pressure boundary,

reactor coolant temperature and pressure, and the corresponding

stresses result in the probabilities of occurrence being

significantly reduced or eliminated, and ultimately result in

minimal consequences. The lone EDG is capable of accepting and

starting required loads within the assumed loading sequence

intervals and continue to operate until offsite power can be

provided to the 480VAC EDS buses. Action statements provide prompt,

specific guidance to ensure sufficiently conservative plant response

should the expected EDG power supply not be available. These action

statements are similar to those in the STS. Therefore, the proposed

license amendment (i.e., changes to 3.7.F.4 and added sections

3.71.5 & 3.7.F.6) does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response

No. The electrical power system specifications support the

equipment required to be operable, commensurate with the current

level of safety, including the equipment requiring an EDG backed

power source. The design review results demonstrate that operation

in the conditions of Cold Shutdown and Refueling, in accordance with

the proposed Technical Specification change, is acceptable from an

accident mitigation standpoint. The basic system functions in Cold

Shutdown and Refueling operating conditions are not changed. One EDG

can supply the necessary electrical power needs during these plant

operating conditions, and in the time frame required to prevent

reactor core/cavity/SFP heatup concerns, with sufficient ``kw

loading'' to spare. The analysis conducted shows that the systems

are capable of performing their design basis functions. Applicable

safety analysis in the Standard Technical Specifications, NUREG

1431, discusses these system requirements as well (i.e., it is not

required to assume a single failure and concurrent loss of all

offsite or all onsite power). Action statements, similar to those in

the Standard Technical Specifications, provide prompt, specific

guidance to ensure sufficiently conservative plant response should

the expected EDG power supply not be available. On this basis, the

proposed license amendment (i.e., changes to 3.7.F.4 and added

sections 3.7.F.5 & 3.7.F.6) does not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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 amendment request: January 29, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to

change the allowable indicated control rod misalignment.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response

No. Based on the Westinghouse evaluation in WCAP-14668, the

Authority has determined that all pertinent licensing basis

acceptance criteria have been met, and the margin of safety as

defined in the TS Bases is not reduced in any of the IP3 licensing

basis accident analysis. Increasing the magnitude of allowed control

rod indicated misalignment (in section 3.10.5) is not a contributor

to the mechanistic cause of an accident evaluated in the FSAR [Final

Safety Analysis Report]. Neither the rod control system nor the rod

position indicator function is being altered. Therefore, the

probability of an accident previously evaluated has not

significantly increased. Because design limitations continue to be

met, and the integrity of the reactor coolant system pressure

boundary is not challenged, the assumptions employed in the

calculation of the offsite radiological doses remain valid.

Therefore, the consequences of an accident previously evaluated

will not be significantly increased.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response

No. Based on the Westinghouse evaluation in WCAP-14668, the

Authority has determined that all pertinent licensing basis

acceptance criteria have been met, and the margin of safety as

defined in the TS Bases is not reduced in any of the IP3 licensing

basis accident analysis. Increasing the magnitude of allowed control

rod indicated misalignment is not a contributor to the mechanistic

cause of any accident. Neither the rod control system nor the rod

position indicator function is being altered. Therefore, an accident

which is new or different than any previously evaluated will not be

created.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response

No. Based on the Westinghouse evaluation in WCAP-14668, the

Authority has

[[Page 19565]]

determined that all pertinent licensing basis acceptance criteria

have been met, and the margin of safety as defined in the TS Bases

is not reduced in any of the IP3 licensing basis accident analysis

based on the changes to safety analyses input parameter values as

discussed in WCAP-14668. Since the evaluations in Section 3.0 of

WCAP-14668 demonstrate that all applicable acceptance criteria

continue to be met, the proposed change will not involve a

significant reduction in margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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 amendment request: January 29, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to

change the allowable indicated control rod misalignment.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response

No. Based on the Westinghouse evaluation in WCAP-14668, the

Authority has determined that all pertinent licensing basis

acceptance criteria have been met, and the margin of safety as

defined in the TS Bases is not reduced in any of the IP3 licensing

basis accident analysis. Increasing the magnitude of allowed control

rod indicated misalignment (in Section 3.10.5) is not a contributor

to the mechanistic cause of an accident evaluated in the FSAR [Final

Safety Analysis Report]. Neither the rod control system nor the rod

position indicator function is being altered. Therefore, the

probability of an accident previously evaluated has not

significantly increased. Because design limitations continue to be

met, and the integrity of the reactor coolant system pressure

boundary is not challenged, the assumptions employed in the

calculation of the offsite radiological doses remain valid.

Therefore, the consequences of an accident previously evaluated

will not be significantly increased.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response

No. Based on the Westinghouse evaluation in WCAP-14668, the

Authority has determined that all pertinent licensing basis

acceptance criteria have been met, and the margin of safety as

defined in the TS Bases is not reduced in any of the IP3 licensing

basis accident analysis. Increasing the magnitude of allowed control

rod indicated misalignment is not a contributor to the mechanistic

cause of any accident. Neither the rod control system nor the rod

position indicator function is being altered. Therefore, an accident

which is new or different than any previously evaluated will not be

created.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response

No. Based on the Westinghouse evaluation in WCAP-14668, the

Authority has determined that all pertinent licensing basis

acceptance criteria have been met, and the margin of safety as

defined in the TS Bases is not reduced in any of the IP3 licensing

basis accident analysis based on the changes to safety analyses

input parameter values as discussed in WCAP-14668. Since the

evaluations in Section 3.0 of WCAP-14668 demonstrate that all

applicable acceptance criteria continue to be met, the proposed

change will not involve a significant reduction in margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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

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

Date of amendment request: March 22, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3.7.1.6, ``Atmospheric Steam Relief

Valves,'' and add a new TS for atmospheric steam relief valve

instrumentation, to ensure that the automatic feature of the steam

generator power-operated relief valve (i.e., atmospheric steam relief

valves) remains operable during Modes 1 and 2. In addition, the

proposed change would add an associated surveillance requiring that a

channel calibration on the steam generator power-operated relief valve

be performed every 18 months.

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 methodologies used in the accident analyses remain

unchanged. The automatic actuation of the Steam Generator Power

Operated Relief Valves is not a new design feature. The effects of

the inadvertent opening of a Steam Generator Power Operated Relief

Valve are currently analyzed as described in Section 15.1.4 of the

Updated Final Safety Analysis Report. The radiological consequences

for the Small Break Loss of Coolant Accident (SBLOCA) event

presented in the Updated Final Safety Analysis Report remain

unchanged. The calculated Peak Clad Temperature is 1849 deg.F

remaining substantially below the 2200 deg.F acceptance limit of 10

CFR 50.46. Although the manual control specification is relocated

from Specification 3.7.1.6 to the new instrumentation specification,

the limiting condition for operation, applicability and action

statements for manual controls remain unchanged. Therefore no

increase in the probability or consequences of any accident

previously evaluated will occur.

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

different kind of accident from any accident previously evaluated?

The automatic actuation of the Steam Generator Power Operated

Relief Valves is not an accident initiator for the SBLOCA event. The

automatic actuation of the Steam Generator Power Operated Relief

Valves currently exists at the South Texas Project and is not a new

design feature. The description of the Steam Generator Power

Operated Relief Valves currently exists in the Updated Final Safety

Analysis Report. This change does not represent a change to the

facility and does not affect the safety functions and reliability of

systems, structures, or components in any new manner. Operating

procedures have a temporary administrative control to ensure the

automatic actuation of the Steam Generator Power Operated Relief

Valves remains operable in Modes 1 and 2. This condition will become

permanent with the approval of this Technical Specification

Amendment proposal. Although the manual control specification is

relocated from Specification 3.7.1.6 to the new instrumentation

specification, the limiting condition for operation, applicability

and

[[Page 19566]]

action statements for manual controls remain unchanged. Since the

automatic actuation of the Steam Generator Power Operated Relief

Valves is not an accident initiator and is not a new design feature

to the facility, no possibility exists for a new or different kind

of accident from those previously evaluated.

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

of safety?

The proposed change results in the calculated Peak Clad

Temperature of 1849 deg.F remaining well below the acceptance limit

of 10 CFR 50.46 and comparable to the results currently described in

the Updated Final Safety Analysis Report. Therefore, the change does

not involve a significant reduction in a margin of safety.

Based on the above, the South Texas Project has evaluated the

proposed Technical Specification change and determined it does not

represent a significant hazards consideration.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,

Sequoyah Nuclear Plant, (SQN), Units 1 and 2, Hamilton County,

Tennessee

Date of application for amendments: March 19, 1999 (TS 99-01).

Brief description of amendments: The proposed amendments would

change the SQN Technical Specifications (TS) for Operating Licenses

DPR-77 (Unit 1) and DPR-79 (Unit 2) by relocating TS Sections 3.8.3.1,

3.8.3.2, and 3.8.3.3 to the SQN Technical Requirements Manual. These

sections provide requirements for electrical overcurrent isolation

devices.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed revision to the TS relocates the requirements for

SQN's electrical equipment protective devices without changing the

current requirements. TVA does not consider these devices to be the

source of any accident; therefore, this administrative relocation of

the requirements will not increase the possibility of an accident.

SQN's electrical equipment protective devices will continue to

provide fault protection for circuits and equipment. Changes to the

relocated requirements will be processed, in accordance with 10 CFR

50.59, to ensure changes are not implemented that would reduce the

functionality or introduce an unreviewed safety question to SQN's

electrical equipment devices. Therefore, the proposed relocation of

the TS requirements for electrical equipment protective devices will

not increase the consequences of an accident.

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

new or different kind of accident from any accident previously

evaluated.

SQN's electrical equipment protective devices ensure proper

operation of plant equipment. These devices are not associated with

accident mitigation or previously evaluated accidents and would not

be the initiator of any new or different kind of accident. The

proposed change does not alter the current functions of these

devices, therefore, this proposed change will not create the

possibility of a new or different kind of accident.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The requirements for SQN's electrical equipment protective

devices are unchanged by the proposed relocation of the requirements

to the SQN Technical Requirements Manual. The function of these

devices and the surveillance testing to ensure operability of these

devices remains unchanged. Any future changes to these requirements

will be evaluated, in accordance with 10 CFR 50.59, to ensure

acceptability and NRC review as required. Accordingly, the proposed

change will not result in a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Section Chief: Sheri R. Peterson.

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.

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: December 16, 1997, as

supplemented August 31, and December 7, 1998.

Brief description of amendment: This amendment changes Technical

Specification 4.7.1.2.1.a.2.a, Auxiliary Feedwater (AFW) System

Surveillance Requirements, by changing the differential pressure and

flow requirements of the steam turbine-driven AFW pump to allow testing

of the pump at a lower speed.

Date of issuance: April 1, 1999.

Effective date: April 1, 1999.

Amendment No.: 87.

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

Technical Specifications.

Date of initial notice in Federal Register: February 11, 1998 (63

FR 6981).

[[Page 19567]]

The August 31, and December 7, 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 April 1, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

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

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: September 1, 1998, as

supplemented on March 19, 1999.

Brief description of amendment: This amendment changes Technical

Specification (TS) \3/4\.9.11, ``Water Level--New and Spent Fuel

Pools,'' and its associated Bases by requiring 23 feet of water above

the top of fuel rods within irradiated fuel assemblies seated in the

storage racks.

Date of issuance: April 8, 1999.

Effective date: April 8, 1999.

Amendment No.: 88.

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

Technical Specifications.

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

FR 50935).

The March 19, 1999, submittal 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 April 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

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

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of application for amendments: November 25, 1998.

Brief description of amendments: The amendments revised the

Technical Specifications (TS) to support on-line replacement of the

Braidwood, Unit 2, vital batteries.

Date of issuance: March 26, 1999.

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

Amendment Nos.: 99 and 99.

Facility Operating License Nos. NPF-72 and NPF-77: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 24, 1999 (64

FR 9185).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wilmington Public Library, 201

S. Kankakee Street, Wilmington, Illinois 60481.

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 application for amendments: January 21, 1999.

Brief description of amendments: The amendments revised the

Technical Specifications (TSs) by relocating TS Section \3/4\.6.I,

``Primary System Boundary-Chemistry'' and associated bases to the

Updated Final Safety Analysis Report (UFSAR) and to applicable plant

procedures.

Date of issuance: March 31, 1999.

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

Amendment Nos.: 187 and 184.

Facility Operating License Nos. DPR-29 and DPR-30: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 24, 1999 (64

FR 9186).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 31, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: January 28, 1999.

Brief description of amendments: The amendments revised Technical

Specifications Section 3.7.13, ``Fuel Handling Ventilation Exhaust

System,'' and associated Bases to correct discrepancies between the

current design and this section.

Date of issuance: March 26, 1999.

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

30 days from the date of issuance.

Amendment Nos.: Unit 1-176; Unit 2-168.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 24, 1999 (64

FR 9187).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: October 15, 1998, as

supplemented December 15, 1998, and January 11 and 21, 1999.

Brief description of amendments: The amendments revised the

Technical Specifications (TSs) to change the heatup, cooldown, and

inservice test limitations for the reactor coolant system of each unit

to a maximum of 26 effective full-power years. The amendments also

revise the TSs for low temperature overpressure protection to reflect

the revised pressure-temperature limits of the reactor vessels.

Date of Issuance: March 30, 1999.

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

90 days from the date of issuance.

Amendment Nos.: Unit 1-302; Unit 2-302; Unit 3-302.

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 2, 1998 (63 FR

66592).

The December 15, 1998, and January 11 and 21, 1999, letters

provided clarifying information that did not change the scope of the

original Federal Register notice and the initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 30, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

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

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

Date of application for amendments: October 15, 1998, as

supplemented

[[Page 19568]]

December 14, 1998, February 18, 1999, and February 23, 1999.

Brief description of amendments: These amendments made several

changes that are administrative in nature. The changes (1) made

editorial changes that delete obsolete material or material adequately

described elsewhere, changed action statement numbers, updated

technical specification (TSs) index pages, and made changes to be

consistent with the guidance provided in the improved standard

technical specifications for Westinghouse reactors (NUREG-1431,

Revision 1); (2) deleted reporting requirements that are duplicated in

various sections of Title 10 of the Code of Federal Regulations; and

(3) relocated the requirement for meteorological monitoring

instrumentation from the TSs to the Licensing Requirements Manual.

The February 18, 1999, and February 23, 1999, letters withdrew a

portion of the amendment request that would have deleted the

description of the site exclusion boundary from the TSs. The

description of the site exclusion boundary will remain in the TS.

Date of issuance: March 26, 1999.

Effective date: Units 1 and 2, as of date of issuance, to be

implemented within 60 days.

Amendment Nos.: 220 and 97.

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

revised the Technical Specifications and licenses.

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

FR 64111).

The December 14, 1998, February 18, 1999, and February 23, 1999,

letters did not change the initial proposed no significant hazards

consideration determination or expand the amendment request beyond the

scope of the initial notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 26, 1999.

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. 50-412, Beaver Valley Power

Station, Unit 2, Shippingport, Pennsylvania

Date of application for amendment: March 10, 1997, as supplemented

July 28, 1997, September 17, 1997, April 30, 1998, January 29, 1999,

and February 26, 1999.

Brief description of amendment: The amendment modifies Technical

Specification 3/4.4.5, ``Steam Generators,'' and its associated Bases

and adds a new license condition to Appendix D to allow repair of steam

generator tubes by installation of sleeves developed by ABB Combustion

Engineering. In addition, the amendment deletes the option for using

the kinetic sleeving methodology previously approved for use at Beaver

Valley Power Station, Unit 2.

Date of issuance: March 26, 1999.

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

days.

Amendment No: 98.

Facility Operating License No. NPF-73. Amendment revised the

Technical Specifications and License.

Date of initial notice in Federal Register: April 23, 1997 (62 FR

19829).

The July 28, 1997, September 17, 1997, April 30, 1998, January 29,

1999, and February 26, 1999, letters provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination or expand the amendment request beyond the

scope of the April 23, 1997, Federal Register notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

Date of amendment request: January 12, 1999, supersedes application

dated May 31, 1996.

Brief description of amendment: The amendment adds an additional

required action to the Limiting Condition for Operation (LCO) 3.9.1,

``Refueling Equipment Interlocks,'' of the RBS Technical

Specifications. The additional action will allow an alternative to the

current action for one or more inoperable refueling equipment

interlocks. The current action is to ``suspend in-vessel fuel movement

with equipment associated with the inoperable interlock(s).'' The

alternative action will be to (1) insert a control rod withdrawal

block, and (2) verify all control rods are fully inserted in core cells

containing one or more fuel assemblies. The amendment also revised the

Bases for LCO 3.9.1 actions to describe the alternative action.

Date of issuance: March 26, 1999.

Effective date: March 26, 1999.

Amendment No.: 104.

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

Technical Specifications.

Date of initial notice in Federal Register: February 10, 1999 (64

FR 6695).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Documents

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

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

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

Date of application for amendment: November 30, 1998.

Brief description of amendment: The amendment changes Technical

Specification 3.1.2, ``Liquid Poison System,'' and its associated Bases

to correct the required concentration and volume of boron solution.

Date of issuance: April 2, 1999.

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

30 days.

Amendment No.: 166.

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

Technical Specifications.

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71970).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 2, 1999.

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.

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

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: April 1, 1998, as supplemented

May 29, June 26, and August 4, 1998.

Brief description of amendment: The amendment revises the Millstone

Unit 3 final safety analysis report (FSAR) by adding a new sump pump

subsystem to address groundwater inleakage through the containment

basemat.

Date of issuance: March 17, 1999.

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

within 60 days from the date of issuance.

[[Page 19569]]

Amendment No.: 168.

Facility Operating License No. NPF-49: Amendment authorized changes

to the FSAR.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19974).

The May 29, June 26, and August 4, 1998, letters provided

clarifying information that did not change the scope of the April 1,

1998, application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment, state

consultation, and final determination of no significant hazards

consideration are contained in a Safety Evaluation dated March 17,

1999.

No significant hazards consideration comments received: No public

comments received. A petition to intervene was received from the

Citizens Regulatory Commission that was dismissed and terminated by the

NRC Atomic Safety Licensing Board (LBP-98-22).

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.

PECO Energy Company, Public Service Electric and Gas Company Delmarva

Power and Light Company, and Atlantic City Electric Company, Docket

Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2

and 3, York County, Pennsylvania

Date of application for amendments: February 4, 1998, as revised

September 29, 1998.

Brief description of amendments: The amendments revise the

Technical Specifications surveillance requirements concerning secondary

containment doors.

Date of issuance: April 7, 1999.

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

within 30 days.

Amendments Nos.: 227 and 230.

Facility Operating License Nos. DPR-44 and DPR-56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 14, 1998 (63 FR

38202).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 7, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Publications

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

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

PA 17105.

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

Generating Station, Salem County, New Jersey

Date of application for amendment: October 22, 1998.

Brief description of amendment: This amendment revises Technical

Specification (TS) 4.8.2.1.b.3 to increase the minimum battery

electrolyte temperature limit from 60 deg.F to 72 deg.F. This change

resolves a discrepancy in the electrolyte temperature assumed in the

Class 1E battery sizing calculations versus the limit specified in the

TSs.

Date of issuance: March 25, 1999.

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

within 60 days.

Amendment No.: 118.

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 2, 1998 (63 FR

66602).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 25, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch

Nuclear Plant, Units 1 and 2, Appling County, Georgia

Date of application for amendments: December 4, 1998.

Brief description of amendments: The amendments make two changes to

the TS. The first change revises the Unit 1 TS Section 2.1.1.2 to

delete the footnote that specifies that the Safety Limit Minimum

Critical Power Ratios are for Cycle 18 only. The second change revises

the TS for both units by deleting Section 5.6.5.b.2) and incorporating

Section 5.6.5.b.1) into Section 5.6.5.b.

Date of issuance: April 1, 1999.

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

30 days from the date of issuance.

Amendment Nos.: Unit 1-215; Unit 2-156.

Facility Operating License Nos. DPR-57 and NPF-5: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 27, 1999 (64 FR

4161).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 1, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424

and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke

County, Georgia

Date of application for amendments: October 15, 1998, as

supplemented by letter dated November 11, 1998.

Brief description of amendments: The amendments change the Vogtle

Electric Generating Plant Unit 1 and 2 Facility Operating Licenses to

delete or modify certain license conditions that have become obsolete

or inappropriate. In addition, the Technical Specifications and Bases

are reissued to reflect new word processing software.

Date of issuance: March 26, 1999.

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

30 days from the date of issuance.

Amendment Nos.: Unit 1-107; Unit 2-85.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments

revised Facility Operating Licenses and the Technical Specifications.

Date of initial notice in Federal Register: December 2, 1998 (63 FR

66602).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia

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

Callaway County, Missouri

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

supplemented by letter dated September 29, 1998, and application dated

July 30, 1998.

Brief description of amendment: The amendment revised Tables 3.3-3,

3.3-4, and 4.3-2 of the technical specifications regarding the

engineered safety feature actuation system (ESFAS) Functional Unit 6.f,

``Loss of Offisite Power--Start Turbine-Driven Pump,'' by establishing

separate requirements for the analog and digital portions of the

associated circuit. The amendment also adds a note to TS Table 4.3-2 to

clarify that the verification of time delays associated

[[Page 19570]]

with ESFAS Functional Units 8.a and 8.b, ``Loss of Power,'' is only

performed as part of the channel calibration.

Date of issuance: April 2, 1999.

Effective date: April 2, 1999, to be implemented within 30 days of

the date of issuance.

Amendment No.: 130.

Facility Operating License No. NPF-30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 16, 1998 (63

FR 69348).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 2, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Elmer Ellis Library,

University of Missouri, Columbia Missouri 65201.

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: November 18, 1998, as

supplemented with additional information by letters dated March 1,

1999, and March 9, 1999.

Brief description of amendment: The amendment revises the pressure/

temperature limits and the low-temperature overpressure protection

requirements in the facility technical specifications.

Date of issuance: April 1, 1999.

Effective date: April 1, 1999.

Amendment No.: 144.

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

Technical Specifications.

Date of initial notice in Federal Register: December 30, 1998.

(63FR71978)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 1, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

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

Dated at Rockville, Maryland, this 14th day of April 1999.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project Management, Office of Nuclear

Reactor Regulation.

[FR Doc. 99-9839 Filed 4-20-99; 8:45 am]

BILLING CODE 7590-01-P