Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating

LicensesInvolving 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 July 11, 1994, through July 22, 1994. The

last biweekly notice was published on July 20, 1994 (59 FR 37060).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, 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 Rules Review and

Directives Branch, Division of Freedom of Information and Publications

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

Washington, DC 20555, 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

20555. The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By September 2, 1994, 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 20555 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, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the

above date. Where petitions are filed during the last 10 days of the

notice period, it is requested that the petitioner promptly so inform

the Commission by a toll-free telephone call to Western Union at 1-

(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to (Project Director): petitioner's name

and telephone number, date petition was mailed, plant name, and

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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 20555, and at the local public document

room for the particular facility involved.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: June 29, 1994

Description of amendment request: The proposed amendment will

delete the requirement to perform alternate train testing to

demonstrate that other, similar, safety-related components are operable

when components are found, or made, inoperable in the safety injection,

residual heat removal, and containment spray systems. The surveillance

requirements, which the licensee refers to as accelerated testing

requirements, affect the following components:

(a)Safety Injection (SI) pumps TS 3.3.1.2.b)

(b) Residual Heat Removal (RHR) Pumps (TS 3.3.1.2.c)

(c) SI and RHR flow paths (TS 3.3.1.2.e)

(d) Containment Spray (CS) (TS 3.3.2.2.a and b)

(e) CS flow paths (TS 3.3.2.2.c)

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment does not involve a significant increase

in the probability of an accident previously evaluated because the

availability of the subject components will not be reduced and the

design and performance of the components are not being changed. The

subject components are provided to mitigate the consequences of

analyzed accidents; therefore their availability has no bearing on

the probability of occurrence of these accidents.

The proposed amendment does not involve a significant increase

in the consequences of an accident previously evaluated. This change

deletes alternate train testing requirements which, if maintained,

could result in loss of the safety function. Elimination of the

requirements will serve to ensure that one train of safety equipment

is always available to mitigate the consequences of an analyzed

accident. The remaining surveillance requirements provide adequate

assurance that the components will be operable when required.

Therefore the consequences of previously evaluated accidents will

not be increase.

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

new or different kind of accident from any previously evaluated

because these proposed changes do not introduce any new modes of

operation or testing, and no physical changes are being made to the

plant. Therefore no new or different kind of accident could be

initiated by this amendment.

3. The proposed revisions do not involve a significant reduction

in the margin of safety since the routine testing requirements that

remain in the Technical Specifications provide adequate assurance

that the components will be operable when needed. Since the

elimination of this accelerated testing will decrease component wear

and improve availability, the margin of safety should be increased.

Since accelerated testing may still occur when component problems

involve a potential common mode failure, margins of safety

associated with the components' abilities to perform their design

functions will not be affected. 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: Hartsville Memorial Library,

Home and Fifth Avenues, Hartsville, South Carolina 29550

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: David B. Matthews

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

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

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of amendment request: June 13, 1994

Description of amendment request: The proposed amendment would make

several changes to the Administrative Controls in Section 6 of

Technical Specifications (TS) for Byron and Braidwood stations. The

proposed changes include: (1) a change to the submittal frequency of

the Radiological Effluent Release Report, (2) a revision to the Shift

Technical Advisor description, (3) clarification of the Shift

Engineer's responsibilities, and (4) editorial changes. The references

to the Semiannual Radiological Effluent Release Report are also revised

in other sections of the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes to Section 6 of Technical Specifications do

not affect any accident initiators or precursors and do not change

or alter the design assumptions for the systems or components used

to mitigate the consequences of an accident.

The proposed changes are administrative in nature and provide

clarification. These changes provide consistency with station

procedures, programs, the Code of Federal Regulations, other

Technical Specifications, and Standard Technical Specifications.

These changes do not impact any accident previously evaluated in the

UFSAR.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not affect the design or operation of

any system, structure, or component in the plant. There are no

changes to parameters governing plant operation; no new or different

type of equipment will be installed. The proposed changes are

considered to be administrative changes. All responsibilities

described in Technical Specifications for management activities will

continue to be performed by qualified individuals.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes do not affect the margin of safety for any

Technical Specification. The initial conditions and methodologies

used in the accident analyses remain unchanged, therefore, accident

analysis results are not impacted.

The proposed changes are administrative in nature and have no

impact on the margin of safety of any Technical Specification. They

do not affect any plant safety parameters or setpoints. The

descriptions for the Shift Technical Advisor and Shift Engineer are

clarified, however, include no reduction to their responsibilities.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Braidwood, the Wilmington Township Public Library, 201 S. Kankakee

Street, Wilmington, Illinois 60481

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

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

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

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of amendment request: July 6, 1994

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3/4.4.5, ``Steam Generators,'' and

the associated bases. Previously the NRC granted amendments to the TSs

which authorized the use of selected steam generator sleeving

processes. In authorizing use of the processes, the amendments cited

references to specific NRC approved vendor technical reports, including

revision number. The proposed changes reference the reports in generic

terms as those that have been approved by the NRC, subject to

limitations and restrictions as noted by the NRC staff. While the

licensee will still have to request NRC approval for application of the

technologies as referenced in vendor reports, the licenses will not

have to be amended each time.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to the Steam Generator section of Technical

Specifications do not affect any accident initiators or precursors

and do not alter the design assumptions for the systems or

components used to mitigate the consequences of an accident. These

changes are editorial changes to the requirements currently

identified in the Technical Specifications. The requirements

approved by the NRC will not be reduced by this request. The

proposed change maintains the administrative controls necessary to

ensure safe plant operation.

The original amendment requested tubesheet sleeves and tube

support plate sleeves as an alternate tube repair method for Bryon

and Braidwood Units 1 and 2. The steam generator sleeves approved

for installation use the Westinghouse process (laser welded joints)

or the Babcock & Wilcox Nuclear Technologies (BWNT) process of

kinetically welded joints. The sleeve configuration was designed and

analyzed in accordance with the criteria of Regulatory Guide (RG)

1.121 and the design requirements of Section III of the American

Society of Mechanical Engineers (ASME) Code. Fatigue and stress

analyses of the sleeved tube assemblies for both processes produced

acceptable results documented in the current Westinghouse and BWNT

Technical Reports. The proposed Technical Specifications change to

allow the use of the current NRC approved laser welded or

kinetically welded sleeving process does not adversely impact any

other previously evaluated design basis accident or the results of

these analyses. Therefore, the editorial changes to the referenced

sleeving Technical Reports will not increase the probability of

occurrence of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes are considered to be administrative

changes. All the requirements described in Technical Specifications

``Acceptance Criteria'' for the Steam Generators will continue to be

implemented as described in the current Technical Reports.

Referencing the current Westinghouse or BWNT Sleeving Technical

Reports currently approved by the NRC and subject to the limitations

and restrictions as noted by the NRC, has no effect upon any design

transient or accident analyses. The proposed changes do not affect

the design or operation of any system, structure, or component in

the plant. There are no changes to parameters governing plant

operation and no new or different type of equipment will be

installed.

The use of the proposed sleeving processes will not introduce

significant or adverse changes to the plant design basis. Stress and

fatigue analyses of the repair have shown the ASME Code and

RG 1.121 allowable values are met. Implementation of the

currently approved laser welded or kinetically welded sleeving will

continue to maintain the overall tube bundle structural integrity at

a level consistent with that of the originally supplied tubing.

Repair of a tube with a sleeve does not provide a mechanism which

would result in an accident outside of the area affected by the

sleeve. Any hypothetical accident as a result of potential tube or

sleeve degradation in the repaired portion of the tube is bounded by

the existing steam generator tube rupture accident analysis. The

tube rupture accident analysis accounts for the installation of

sleeves and the impact on current plugging level analyses. The

sleeve design does not affect any other component or location on the

tube outside of the immediate area repaired.

Thus, the possibility of a new or different type of accident

from any accident previously evaluated is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change is administrative in nature and has no

impact on the margin of safety of any Technical Specification.

Specific technical reports are no longer referenced in Technical

Specifications. An editorial change is made to TS referencing the

current NRC approved vendor Technical Report, subject to the

limitations and restrictions noted by the NRC. The initial

conditions and methodologies used in the accident analyses remain

unchanged.

The laser welded and kinetically welded sleeving repair of

degraded steam generator tubes has been shown by analysis to restore

the integrity of the tube bundle to its original design basis

condition. The safety factors used in the design of sleeves for the

repair of degraded tubes are consistent with the safety factors in

the ASME Boiler and Pressure Vessel Code used in steam generator

design. The design of the tube sleeves has been verified by testing

to preclude leakage during normal and postulated accident

conditions. Installation of either type of vendor sleeve using the

current approved process will continue to maintain the structural

integrity of the steam generator tubes.

Thus, these changes do not involve a significant reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Braidwood, the Wilmington Township Public Library, 201 S. Kankakee

Street, Wilmington, Illinois 60481

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

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

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

Date of amendment request: June 9, 1994

Description of amendment request: The proposed amendments would

revise the LaSalle County Station, Units 1 and 2, Technical

Specifications (TS), Appendix A, in order to facilitate implementation

of the Thermal Limits portion of the General Electric Average Power

Range Monitor (APRM)/Rod Block Monitor (RBM)/TS Improvement Program

(ARTS).

Specifically, the proposed TS change will create power and flow

dependent Minimum Critical Power Ratio (MCPR) and Maximum Average

Planar Linear Heat Generation Rate (MAPLHGR) limits, and other

administrative changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because:

The probability of an accident previously evaluated will not

increase as a result of this change, because no changes to plant

systems will occur. All changes are related to core monitoring

software, and there will be no physical changes to equipment.

The consequences of an accident previously evaluated will not

increase as a result of the proposed changes. The power- and flow-

dependent MCPR and MAPLHGR limits incorporate sufficient

conservatism so the safety limit MCPR [SLMCPR] (operating limit MCPR

[OLMCPR] for automatic flow control) and the fuel thermal-mechanical

limits will not be violated for any power and flow condition.

Because these limits are protected during normal operation, the

consequences of any transient will not increase with this change in

limit definition. General Electric has verified in Attachment E that

the introduction of Arts will not cause any change in the Licensing

Basis PCT [Peak Centerline Temperature] resulting from a Loss-Of-

Coolant Accident [LOCA], nor any change in the results satisfying

the other LOCA acceptance criteria of 10 CFR 50.46 and Section

15.6.5 of NUREG-0800 (Standard Review Plan), which are: cladding

oxidation, metal-water reaction (hydrogen generation), coolable

geometry and long-term cooling.

The proposed changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated

because:

Since no physical changes to any plant system are occurring,

there will be no new or different types of accidents created by this

change. No interactions between equipment systems will be changed in

any manner.

The proposed changes do not involve a significant reduction in a

margin of safety because:

The power- and flow-dependent MCPR and MAPLHGR limits will

sufficiently protect the SLMCPR (OLMCPR for automatic flow control)

and the fuel thermal-mechanical limits at all power and flow

conditions. The ARTS limits conservatively assure that all licensing

criteria are satisfied without setdown of the flow referenced APRM

scram and rod block trips. The limits were developed using NRC

approved methods, and satisfy the same NRC approved criteria that

the APRM setdown requirement does.

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: Public Library of Illinois

Valley Community College, Rural Route No. 1, Ogelsby, Illinois 61348

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

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

Date of amendment request: July 13, 1994

Description of amendment request: The requested amendments would

allow the testing interval in Technical Specification Surveillance

Requirement 4.6.2 for the air or smoke flow test through each

containment spray header to be increased from 5 to 10 years. The

licensee states that the proposed amendments are consistent with NRC

staff guidance contained in NUREG-1366, ``Improvements to Technical

Specifications Surveillance Requirements,'' and Generic Letter 93-05,

``Line-Item Technical Specifications Improvements to Reduce

Surveillance Requirements for Testing During Power Operation.'' In

addition, the amendments would also remove an obsolete footnote related

to the Catawba Unit 1 first refueling. The licensee's application

jointly addressed both its Catawba and McGuire Nuclear Stations. This

notice addresses those aspects applicable only to the Catawba Station.

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:

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Increasing the surveillance interval of TS [Technical

Specification] 4.6.2d from five to ten years will have no impact

upon the probability of any accident, since the NS [containment

spray] system is not accident initiating equipment. Also, since

Catawba's... flow test histor[y] support[s] making the proposed

change, system response following an accident will not be adversely

affected. Therefore, the requested amendments will not result in

increased accident consequences. Deletion of the obsolete footnote

as indicated in the Catawba TS markup is purely an administrative

change, and therefore will have no impact upon either the

probability or consequences of any accident.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. As stated above, the NS system is not accident initiating

equipment. No new failure modes can be created from an accident

standpoint. The plant will not be operated in a different manner.

Deletion of the Catawba obsolete footnote has no bearing on any

accident initiating mechanisms.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. Plant safety margins will be

unaffected by the proposed changes. The NS system will still be

capable of fulfilling its required safety function, since plant

operating experience supports the proposed change. Finally, the

proposed amendments are consistent with the NRC position and

guidance set forth in NUREG-1366 and Generic Letter 93-05. Deletion

of the Catawba obsolete footnote will not result in any impact to

plant safety margins.

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments do not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

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

Date of amendment request: July 13, 1994

Description of amendment request: The requested amendments would

allow the testing interval in Technical Specification Surveillance

Requirement 4.6.2 for the air or smoke flow test through each

containment spray header to be increased from 5 to 10 years. The

licensee states that the proposed amendments are consistent with NRC

staff guidance contained in NUREG-1366, ``Improvements to Technical

Specifications Surveillance Requirements,'' and Generic Letter 93-05,

``Line-Item Technical Specifications Improvements to Reduce

Surveillance Requirements for Testing During Power Operation.'' The

licensee's application jointly addressed both its Catawba and McGuire

Nuclear Stations. This notice addresses those aspects applicable only

to the McGuire Station.

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:

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Increasing the surveillance interval of TS [Technical

Specification] 4.6.2d from five to ten years will have no impact

upon the probability of any accident, since the NS [containment

spray] system is not accident initiating equipment. Also, since...

McGuire's flow test histor[y] support[s] making the proposed change,

system response following an accident will not be adversely

affected. Therefore, the requested amendments will not result in

increased accident consequences. ...

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. As stated above, the NS system is not accident initiating

equipment. No new failure modes can be created from an accident

standpoint. The plant will not be operated in a different manner.

...

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. Plant safety margins will be

unaffected by the proposed changes. The NS system will still be

capable of fulfilling its required safety function, since plant

operating experiences supports the proposed change. Finally, the

proposed amendments are consistent with the NRC position and

guidance set forth in NUREG-1366 and Generic Letter 93-05. ...

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments do not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee

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

Date of amendment request: December 8, 1993, as supplemented April

20, 1994.

Description of amendment request: The amendments would revise

Technical Specification (TS) 3.4 to address the need to bypass

automatic initiation of the Emergency Feedwater (EFW) system when the

main feedwater pump discharge pressure is below actuation setpoint

during startup and shutdown in order to prevent inadvertent actuation.

The proposed amendment is in response to NRC Inspection Report 50-269,

50-270, 50-287/90-30 (Inspector Followup Item 90-30-02), which

determined that the existing TSs regarding initiation circuitry for the

EFW system were inadequate. The amendments would also delete

operability requirements for the Emergency Condenser Cooling Water

(ECCW) system. The licensee determined that the ECCW system is not

required to remove decay heat following any design basis event.

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 amendment request would not] involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Each accident analysis addressed within the Oconee FSAR [Final

Safety Analysis Report] has been examined with respect to changes

proposed within this amendment request. The design basis of the

Emergency Feedwater (EFW) System is to supply feedwater to the steam

generators in the event Main Feedwater is lost. The EFW system

provides the required flow rate to cool the RCS [reactor coolant

system] down to the point at which the Decay Heat Removal System is

designed to operate. The EFW system is also designed to cool the RCS

following a small break LOCA [loss of coolant accident]. Changes

included within this amendment request are provided to clarify

requirements for the operability of EFW. Specifically, theses

changes clarify that automatic initiation circuitry due to low main

feedwater pump discharge pressure or low hydraulic oil pressure may

be bypassed when the reactor is shutdown to prevent inadvertent

actuation. In addition, these changes provide that if an EFW pump is

inoperable due only to the inoperability of automatic initiation,

cooldown to below 250 deg.F is not necessary after the reactor is

shutdown. Accident analysis for the loss of main feedwater, and

subsequent initiation of EFW, assumes initial conditions of the

reactor at full power operation. The utilization of criticality for

this specific automatic initiation circuitry to be operable ensures

the EFW system is operated within the boundaries of design basis for

Oconee while also providing a reasonable margin to prevent

inadvertent actuation. It is not possible to place this automatic

initiation circuitry in service prior to exceeding 250 deg.F because

the main feedwater pump discharge pressure is well below the

initiation setpoint at this value. Manual initiation circuitry

operability is required prior to exceeding an RCS temperature of

250 deg.F. This change only clarifies existing configuration and

control for the Oconee units and does not increase the probability

or consequences of any accident previously evaluated.

This change also removes the requirement for Emergency Condenser

Cooling Water (ECCW) System operability for the removal of decay

heat using the secondary systems. The ability to provide flow

through the condenser from the ECCW system is a preferred method for

decay heat removal. However, this mode of operation is not necessary

to prevent or mitigate any accident previously evaluated. The

primary success path for decay heat removal following loss of

station power events, and thus loss of normal CCW flow, is the use

of the turbine driven EFW pump providing flow to the steam

generators and heat removal via the main steam safety relief valves

to the atmosphere. Analysis has shown that sufficient inventory

exists in secondary systems, as limited by Technical Specification

3.4.4, to provide for decay heat removal.

Therefore, this proposed change deletes the requirement for ECCW

for secondary systems decay heat removal. The probability or

consequences of any design basis accident are not increased by this

change. As such, this change will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

(2) [The amendment request would not] create the possibility of

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

previously evaluated.

Changes included within this amendment request are provided to

clarify existing requirements for operability of the EFW System and

remove the requirement for ECCW flow through the condenser for decay

heat removal. Operation of Oconee units in accordance with these

Technical Specifications will not create any failure modes not

bounded by previously evaluated accidents. Previously evaluated

accidents assume an initial condition of power operation for loss of

main feedwater events. Providing for automatic initiation prior to

criticality ensures operation within the bounds of design analysis.

Previously evaluated accidents also assume the removal of decay

heat, following loss of normal CCW flow, to be via the main steam

safety relief valves to the atmosphere which eliminates the need for

ECCW operability. Consequently, this change will not create the

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

accident previously evaluated.

(3) [The amendment request would not] involve a significant

reduction in a margin of safety.

The design basis of the EFW system is to supply feedwater to the

steam generators in the event Main Feedwater is lost. By providing

clarification that manual initiation circuitry is operable prior to

exceeding an RCS temperature of 250 deg.F and automatic initiation

circuitry, due to low main feedwater discharge pressure or low

hydraulic oil pressure, is operable prior to criticality, there is

no significant reduction in the margin of safety associated with

this amendment request. The ECCW system is designed to provide a

means to remove decay heat without a loss of secondary side

inventory. However, analysis has shown that sufficient secondary

side inventory exist, as specified by Technical Specification 3.4.4,

to provide for coping with loss of station power events.

Furthermore, even though this method of decay heat removal is

desirable, Oconee PRA [probabilistic risk assessment] studies do not

model the loss of ECCW for accident precursors since it is not

required and margins of safety are not reduced if it is not

available. Changes included within this amendment request clarify

existing requirements for the operability of secondary system for

decay heat removal based on previously evaluated accidents. As such,

all margins of safety are preserved. Therefore, there will be no

reduction in any margin of safety.

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

this review, it appears that the three standards of 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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

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: June 2, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) sections 3.4.6.1 and 3.4.6.2

related to reactor coolant system (RCS) operational leakage and leakage

detection instrumentation. The proposed amendment would revise the TSs

to be in accordance with the standard TSs in NUREG-1431 in so far as

the plant-specific design will allow. The proposed changes relate to

the limiting conditions for operation and the surveillance requirements

for the four primary instruments used to detect RCS leakage. Changes

are also proposed for the index and definition sections. A new TS,

section 3/4.5.4, is proposed to address reactor coolant pump seal

injection flow.

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 probability of occurrence of a previously evaluated

accident, i.e., loss of coolant accident (LOCA), is not increased

because the ability of the plant operators to detect RCS leakage and

take appropriate corrective action is not changed. The proposed

change will continue to ensure that diverse means for detecting

extremely small leaks are available to plant operators. In

addition, the proposed amendment does not change the operational

leakage limits. The seal injection flow limit is not affected by

this proposed change. Due to these three factors, the probability of

occurrence of a LOCA is not increased. The consequences of an

accident previously evaluated are not significantly increased

because the proposed changes do not affect the ability of the

various safety systems to perform their intended function. The

leakage detection monitors do not initiate any automatic function to

mitigate the consequences of a LOCA. They provide an early

indication of RCS leakage. The operational leakage limits are not

affected by this proposed change and they do not initiate any

automatic function to mitigate the consequences of a LOCA. The

proposed change to the seal injection flow requirement will continue

to ensure that ECCS flow will be as assumed in the accident

analyses.

Therefore, based on the continued ability of the leakage

detection monitors and independent monitoring capabilities to detect

extremely small leaks, the fact that this proposed amendment does

not change the operational leakage limits, the seal injection flow

limit is not affected by this proposed change, and that the proposed

changes do not affect the ability of the various safety systems to

perform their intended functions, this proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.2. Does the change create the

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

previously evaluated?

The proposed amendment does not change the plant configuration

in a way which introduces a new potential hazard to the plant. Since

design requirement[s] continue to be met and the integrity of the

RCS pressure boundary is not challenged, no new failure mode has

been created. As a result, an accident which is different than any

already evaluated in the Updated Final Safety Analysis Report

(UFSAR) will not be created due to this change.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

of safety?

The proposed change does not involve a significant reduction in

a margin of safety since the operational leakage limits will not be

affected. Continued plant operation will not be permitted if

operational leakage exceeds the current technical specification

limits. The operational leakage limits establish limits which ensure

that any RCS leakage does not compromise safety. The protection of

the RCS pressure boundary from degradation and the core form [from]

inadequate cooling, in addition to preventing the accident analyses

radiation release assumptions from being exceeded, is the main

purpose of the operational leakage limits. The ability to detect and

quantify operational leakage allows plant operators to perform

actions to place the plant in a safe condition when leakage rate

indicates possible RCS pressure boundary degradation. The proposed

change will continue to ensure that diverse measurement means are

available to provide the plant operators with an early indication of

extremely small RCS leakage. Therefore, [the change is] allowing

action to be taken to place the plant in a safe condition when RCS

leakage indicates possible RCS pressure boundary leakage.

The proposed addition of the separate seal injection

specification will not change the flow limit on seal injection. The

new specification will continue to ensure that seal injection flow

is limited. This will ensure that sufficient flow to the reactor

core is provided during accident conditions. The proposed

elimination of the Mode 4 applicability, for seal injection flow

specification, will not involve a significant reduction in the

margin of safety since high seal injection flow is less critical as

a result of the lower initial RCS pressure and decay heat removal

requirements in Mode 4.

Therefore, this proposed change does not involve a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 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, Pennsylvania 15001.

Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,

Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Walter R. Butler

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 6, 1994

Description of amendment request: The proposed admendment would

revise the technical specifications (TSs) by relocating the seismic and

meteorological monitoring instrumentation and their associated

requirements from the TSs to the Waterford 3 updated final safety

analysis report and plant procedures pursuant to the NRC final policy

statement on TSs improvements for nuclear power reactors. The final

policy statement was published in the Federal Register on Thursday,

July 22, 1993.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change relocates Seismic and Meteorological

Monitoring Instrumentation requirements from the TS to licensee

controlled documents consistent with the NRC Policy Statement on

Technical Specification Improvements. Criterion 1 of the Policy

Statement indicates that the TS should include installed

instrumentation that is used to detect, and indicate in the control

room, a significant abnormal degradation of the reactor coolant

pressure boundary. This criterion is intended to ensure that the TS

control those instruments specifically installed to detect excessive

reactor coolant system leakage. This criterion is not interpreted to

include instrumentation used to detect precursors to reactor coolant

pressure boundary leakage (e.g., loose parts monitor, seismic

instrumentation, valve position indicators). Combustion Engineering

and the NRC have previously determined that relocating Seismic and

Meteorological Monitoring Instrumentation requirements from the TS

does not affect any material condition of the plant that could

directly contribute to causing or mitigating the effects of an

accident.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

The proposed change will not involve any design change or

modification to the plant. The proposed change will not alter the

operation of the plant or the manner in which it is operated. Any

subsequent change to the Seismic or Meteorological Monitoring

Instrumentation requirements will undergo a review in accordance

with the criteria of 10 CFR 50.59 to ensure that the change does not

involve an unreviewed safety question.

Therefore, the proposed change will not create the possibility

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

evaluated.

The proposed change will relocate Seismic and Meteorological

Monitoring Instrumentation requirements from the TS to licensee

controlled documents subject to the criteria of 10 CFR 50.59. The

proposed change will have no adverse impact on any protective

boundary or safety limit.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 22, 1994

Description of amendment request: The proposed amendment would

revise the technical specifications (TSs) to change three plant

protection system (PPS) trip setpoints to be consistent with the

current setpoint/uncertainty methodology being implemented at Waterford

3. The change adjusts the affected TSs values in a more conservative

direction.

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:

Implementing the proposed change will not affect any design

basis accident. The revised trip and actuation setpoints are based

upon the same analytical limits that form the basis for the current

trip and actuation setpoints. The design basis for each trip and

actuation setpoint was verified to be consistent with the

appropriate accident analyses as part of the process of revising the

PPS setpoint analysis. The proposed changes in trip and actuation

setpoints are all in the conservative (away from the analytical

limits) direction. Therefore, the proposed change will not involve a

significant increase in the probability or consequences of any

previously analyzed accident.

Plant operation and the manner in which the plant is operated

will not be altered as a result of implementing the proposed change

since no new system or design change is being implemented.

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

new or different kind of accident from any accident previously

evaluated.

The current safety margins of the affected trip setpoints and

allowable values is preserved by the proposed change. This is

assured by retaining the current analytical limit for the affected

parameters. Since the analytical limits are not affected and the

total channel uncertainty is increased, the margin of safety for the

affected trip setpoints and allowable values is preserved.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 22, 1994

Description of amendment request: The proposed amendment would

revise the technical specifications (TSs) to replace the generic

control room outside air intake (CROAI) radiation alarm/trip setpoint

(less than or equal to 2x background) with a specific setpoint (less

than or equal to 4.09E-5). The new setpoint is based on radioactive

material concentrations in the control room not exceeding the derived

air concentrations (DAC) occupational values listed in 10 CFR Part 20,

Appendix B, Table 1, Column 3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change replaces the current CROAI radiation monitor

alarm/trip setpoint of less than or equal to 2x background with a

fixed value independent of background radiation. The new setpoint

will continue to provide protection to plant personnel such that

occupational radiation exposure is maintained within the limits of

10 CFR 20 during normal plant operation, anticipated operational

occurrences or design basis accidents.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

The proposed change will replace the generic CROAI radiation

monitor alarm/trip setpoint with a setpoint derived from a site-

specific calculation. The proposed change will not alter the

operation of the plant or the manner in which it is operated.

Therefore, the proposed change will not create the possibility

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

evaluated.

The proposed change will replace the current CROAI radiation

monitor setpoint with a new setpoint that will ensure occupational

radiation exposure will not exceed the DAC limits of 10 CFR 20. The

proposed change has no adverse impact on protective boundaries,

safety limits, or margin of safety.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

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

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

Date of amendment request: July 19, 1994

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications by relocating

cycle-specific parameter limits from the Technical Specifications to

the Core Operating Limits Report (COLR). Presently, the parameter

limits for Turkey Point Units 3 and 4 are calculated using NRC-approved

methodologies. These limits are evaluated for every reload cycle and

may be revised by a license amendment as appropriate, to reflect

changes to cycle-specific variables.The curves to be relocated include

(a) TS Figure 3.1-2, Rod Bank Insertion Limits versus Thermal Power

curve, and (b) TS Figure 3.2-2, K(Z) Normalized FQ(Z) as a

Function of Core Height curve.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The removal of cycle-specific Rod Bank Insertion limits and the

K(Z) curve from the Turkey Point Units 3 and 4 Technical

Specifications is administrative in nature and has no impact on the

probability or consequences of any Design Bases Event (DBE)

occurrences which was previously evaluated. The determination of the

Rod Bank Insertion limits and K(Z) curve will be performed using

methodology approved by the NRC and poses no significant increase in

the probability or consequences of any accident previously

evaluated.

The Rod Bank Insertion limits and K(Z) curve will be evaluated

every cycle to ensure proper compliance with the Updated Final

Safety Analysis Report (UFSAR). These limits will be evaluated in

accordance with 10 CFR Sec. 50.59, which ensures that the reload

will not involve an increase in the probability of occurrences or

consequences of an accident previously evaluated. 10 CFR Sec. 50.59

(2) states that a proposed change involves an unreviewed safety

question (i) if the probability of occurrence or the consequences of

an accident or malfunction of equipment important to safety

previously evaluated in the safety analysis report may be increased.

Consequently, since any change to the reload core design analysis

must be evaluated relative to the more restrictive evaluation

criterion of 10 CFR Sec. 50.59, then operation of the facility in

accordance with the proposed amendments would not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The removal of the Rod Bank Insertion limits and K(Z) curve from

the Technical Specifications is administrative in nature and has no

impact, nor does it contribute in any way to the possibility of a

new or different kind of accident from any accident previously

evaluated. No new accident scenarios, failure mechanisms or limiting

single failure events are introduced as a result of the proposed

change.

The generation of the Rod Bank Insertion limits and K(Z) curve

will be performed using NRC-approved methodology and are submitted

to the NRC, as a revision to the COLR, to allow the NRC staff to

trend. The Technical Specifications will continue to require

operation within the core operating limits and appropriate actions

will be taken if these limits are exceeded.

10 CFR Sec. 50.59 permits a licensee to make changes in the

facility as described in the safety analysis report without prior

Commission approval, provided that the proposed changes does not

involve an unreviewed safety question. 10 CFR Sec. 50.59 (2) states

that a proposed change involves an unreviewed safety question (ii)

if a possibility for an accident or malfunction of a different type

than any evaluated previously in the safety analysis report may be

created. Consequently, since any change to the reload core design

analysis must be evaluated relative to the more restrictive

evaluation criterion of 10 CFR Sec. 50.59, then operation of the

facility in accordance with the proposed amendments would not create

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

accident previously evaluated.

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

amendments would not involve a significant reduction in a margin of

safety.

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

Bank Insertion limits and K(Z) curve from the Technical

Specifications. The methodology for the reload core design analysis

have been approved by the NRC and does not constitute a significant

reduction in the margin of safety.

The supporting Technical Specification values are defined by the

accident analyses which are performed to conservatively bound the

operating conditions defined by the Technical Specifications. The

development of the limits for future reloads will continue to

conform to the methodology described in NRC approved documentation.

In addition, each future reload will involve a 10 CFR 50.59 review

to assure that operation of the units within the cycle specific

limits will not involve a reduction in a margin of safety. 10 CFR

Sec. 50.59 (2) states that a proposed change involves an unreviewed

safety question (iii) if the margin of safety as defined in the

basis for any technical specification is reduced. Consequently,

since any change to the reload core design analysis must be

evaluated relative to the more restrictive evaluation criterion of

10 CFR Sec. 50.59, then operation of the facility in accordance with

the proposed amendments would not involve a significant reduction in

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,

P.C., 1615 L Street, NW., Washington, DC 20036

NRC Project Director: Victor M. McCree, Acting

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

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

Date of amendment request: July 19, 1994

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications and its associated

BASES, which address the maximum allowed reactor thermal power

operation with inoperable main steam safety valves (MSSVs).

Westinghouse issued Nuclear Safety Advisory Letter (NSAL) 94-001 which

notified the licensee of a deficiency in the basis of the Turkey Point

Technical Specification 3/4.7.1, which allows the plant to operate at

reduced power levels with a specified number of MSSVs inoperable. This

amendment request corrects the allowable power level with inoperable

MSSVs.

The licensee also proposed changes to the TS 3.7.1.1 applicability

statement to indicate that, for mode 3 only, the actions are required

when the Reactor Trip System breakers are in the closed position and

the Control Rod Drive System is capable of rod withdrawal.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The new power range neutron flux high setpoint values will ensure

that the secondary side steam pressure will remain below 110 percent

of the design value following a Loss of Load/Turbine Trip event,

when one or more main steam safety valves (MSSVs) are declared

inoperable. The proposed change will not impact the classification

of the Loss of Load/Turbine Trip event as a Condition II probability

event (faults of moderate frequency) per ANSI - N18.2, 1973.

Accordingly, since the new power range neutron flux setpoints will

maintain the capability of the MSSVs to perform their pressure

relief function associated with a Loss of Load/Turbine Trip event,

there will be no effect on the probability or consequences of an

accident previously evaluated.

In addition, the proposed change to the applicability statement

of TS 3.7.1.1, will not effect the probability or consequences of an

accident previously evaluated, since the proposed plant condition

with the reactor trip breakers open and the rod control system not

capable of withdrawing rods is an analyzed safe shutdown condition.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes do not involve any change to the configuration

or method of operation of any plant equipment, and no new failure

modes have been defined for any plant system or component. The new

power range neutron flux high setpoints will maintain the capability

of the MSSVs to perform their pressure relief function to ensure the

secondary side steam design pressure is not exceeded following a

Loss of Load/Turbine Trip event. Therefore, since the function of

the MSSVs is unaffected by the proposed changes, the possibility of

a new or different kind of accident from any accident previously

evaluated is not created.

In addition, the proposed change to the applicability statement

of TS 3.7.1.1, will not create the possibility of a new or different

kind of accident from any accident previously evaluated, since the

proposed plant condition with the reactor trip breakers open and the

rod control system not capable of withdrawing rods is an analyzed

safe shutdown condition.

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

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes to the Technical Specifications [do] not

involve a significant reduction in a margin of safety. The algorithm

methodology used to calculate the new power range neutron flux high

setpoints is conservative and bounding since it is based on a number

of inoperable MSSVs per loop; i.e., if only one MSSV in one loop is

out of service, the applicable power range setpoint would be the

same as if one MSSV in each loop were out of service. Another

conservatism with the algorithm methodology is with the assumed

minimum total steam flow rate capability of the operable MSSVs. The

assumption is that if one or more MSSVs are inoperable per loop, the

inoperable MSSVs are the largest capacity MSSVs, regardless of which

capacity MSSVs are actually inoperable. Therefore, since the power

range neutron flux setpoints calculated for the proposed changes

using the algorithm methodology are more conservative and ensure the

secondary side steam design pressure is not exceeded following a

Loss of Load/Turbine Trip event, this proposed license amendment

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

In addition, the proposed change to the applicability statement

of TS 3.7.1.1, does not involve a significant reduction in a margin

of safety, since the proposed plant condition with the reactor trip

breakers open and the rod control system not capable of withdrawing

rods is an analyzed safe shutdown condition.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,

P.C., 1615 L Street, NW., Washington, DC 20036

NRC Project Director: Victor M. McCree, Acting

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

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

Date of amendment request: July 19, 1994

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications (TS) by revising

Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f., to delete the

specific reference in the TS of the American Society for Testing and

Materials (ASTM) testing standard being used to meet TS testing

requirements. The Emergency Diesel Generator (EDG) fuel oil TS

Surveillance Requirements will be replaced with a requirement to test

the EDG fuel oil in accordance with the Turkey Point Units 3 and 4

Diesel Fuel Oil Testing Program.

The licensee proposes the addition of ACTION statements g. and h.

of TS 3.8.1.1 to address the required action in the event the diesel

fuel oil does not meet the Diesel Fuel Oil Testing Program 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) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes to the Technical Specifications will permit

the Technical Specification required testing of Emergency Diesel

Generator (EDG) fuel oil in accordance with the Turkey Point Units 3

and 4 Diesel Fuel Oil Testing Program. The proposed change will

permit FPL to use more recent editions of the American Society for

Testing and Materials (ASTM) standards currently listed in Technical

Specification Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f.

Prior to changing the Diesel Fuel Oil Testing Program, the proposed

change will be evaluated pursuant to Title 10 Code of Federal

Regulations Sec. 50.59 (10 CFR Sec. 50.59), ``Changes, tests, and

experiments.'' Title 10 CFR Sec. 50.59 permits a licensee to make

changes in the procedures as described in the safety analysis report

without prior Commission approval, provided that the proposed

changes [do] not involve an unreviewed safety question.

Title 10 CFR Sec. 50.59(a)(2) states that a proposed change

involves an unreviewed safety question (i) if the probability of

occurrence or the consequences of an accident or malfunction of

equipment important to safety previously evaluated in the safety

analysis report may be increased. Consequently, since any change to

the Diesel Fuel Oil Testing Program, including the ASTM standard or

ASTM edition standard to be used to evaluate EDG fuel oil

acceptability, the change must be evaluated relative to the more

restrictive evaluation criterion of 10 CFR Sec. 50.59, then

operation of the facility in accordance with the proposed amendments

would not involve a significant increase in the probability or

consequences of an accident previously evaluated. The EDG fuel oil

TS Surveillance Requirements will be replaced with a requirement to

test the EDG fuel oil in accordance with the Turkey Point Units 3

and 4 Diesel Fuel Oil Testing Program.

ACTION statement g. of TS 3.8.1.1 is added to address the

required action in the event the new fuel oil properties do not meet

the Diesel Fuel Oil Testing Program limits. A failure to meet the

API gravity, kinematic viscosity, flash point or clarity limits is

cause for rejecting the new fuel oil prior to the addition to the

Diesel Fuel Oil Storage Tanks, but does not represent a failure to

meet the Limiting Condition for Operation (LCO) of TS 3.8.1.1, since

the new fuel oil has not been added to the storage tanks. Provided

these new fuel oil properties are met subsequent to the addition of

the new fuel oil to the storage tanks, 30 days is provided to

complete the analyses of the other fuel oil properties specified in

Table 1 of ASTM-D975-81, except sulfur which may be performed in

accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the

other new fuel oil properties specified in Table 1 of ASTM-D975-81

are not met, ACTION statement g. of TS 3.8.1.1 provides an

additional 30 days to meet the Diesel Fuel Oil Testing Program

limits. This additional 30 day period is acceptable because the fuel

oil properties of interest, even if they are not within limits,

would not have an immediate effect on EDG operation.

ACTION statement h. of TS 3.8.1.1 is added to address the

required action in the event the stored fuel oil total particulates

does not meet the Diesel Fuel Oil Testing Program limits. Fuel oil

degradation during long term storage shows up as an increase in

particulate, due mostly to oxidation. The presence of particulate

does not mean the fuel oil will not burn properly in a diesel

engine. The frequency for performing surveillance on stored fuel oil

is based on stored fuel oil degradation trends which indicate that

particulate concentration is unlikely to change significantly

between surveillances.

Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil

Testing Program, FPL will need to determine if the proposed program

change is at least as, if not more, effective, in detecting

unsatisfactory fuel oil. The EDGs will thus continue to function as

designed and the probability or consequences of previously evaluated

accidents will be unaffected.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed changes to the Technical Specifications will permit

the Technical Specification required testing of Emergency Diesel

Generator fuel oil using more recent editions of the American

Society for Testing and Materials standards listed in Technical

Specification Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f.

Prior to changing the edition of the previously approved ASTM

standard being used to evaluate the EDG fuel oil, the proposed

edition standard will be evaluated pursuant to 10 CFR Sec. 50.59,

``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits

a licensee to make changes in the procedures as described in the

safety analysis report without prior Commission approval, provided

that the proposed changes does not involve an unreviewed safety

question. Title 10 CFR Sec. 50.59(a)(2) states that a proposed

change involves an unreviewed safety question (ii) if a possibility

for an accident or malfunction of a different type than any

evaluated previously in the safety analysis report may be created.

Consequently, since any change to the edition of the ASTM standard

to be used to evaluate EDG fuel oil acceptability must be evaluated

relative to the more restrictive evaluation criterion of 10 CFR

Sec. 50.59, then operation of the facility in accordance with the

proposed amendments would not create the possibility of a new or

different kind of accident from any accident previously evaluated.

ACTION statement g. of TS 3.8.1.1 is added to address the

required action in the event the new fuel oil properties do not meet

the Diesel Fuel Oil Testing Program limits. A failure to meet the

API gravity, kinematic viscosity, flash point or clarity limits is

cause for rejecting the new fuel oil prior to the addition to the

Diesel Fuel Oil Storage Tanks, but does not represent a failure to

meet the Limiting Condition for Operation (LCO) of TS 3.8.1.1, since

the new fuel oil has not been added to the storage tanks. Provided

these new fuel oil properties are met subsequent to the addition of

the new fuel oil to the storage tanks, 30 days is provided to

complete the analyses of the other fuel oil properties specified in

Table 1 of ASTM-D975-81, except sulfur which may be performed in

accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the

other new fuel oil properties specified in Table 1 of ASTM-D975-81

are not met, ACTION statement g. of TS 3.8.1.1 provides an

additional 30 days to meet the Diesel Fuel Oil Testing Program

limits. This additional 30 day period is acceptable because the fuel

oil properties of interest, even if they are not within limits,

would not have an immediate effect on EDG operation.

ACTION statement h. of TS 3.8.1.1 is added to address the

required action in the event the stored fuel oil total particulates

[do] not meet the Diesel Fuel Oil Testing Program limits. Fuel oil

degradation during long term storage shows up as an increase in

particulate, due mostly to oxidation. The presence of particulate

does not mean the fuel oil will not burn properly in a diesel

engine. The frequency for performing surveillance on stored fuel oil

is based on stored fuel oil degradation trends which indicate that

particulate concentration is unlikely to change significantly

between surveillances.

Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil

Testing Program, FPL will need to determine if the proposed program

change is at least as, if not more, effective, in detecting

unsatisfactory fuel oil. Since the proposed changes do not involve a

change in the design of any plant system or component, and since the

proposed changes will need to evaluate the effect of any ASTM

standard edition change on the level of EDG reliability, the change

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

of accident from any accident previously evaluated.

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

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes to the Technical Specifications will permit

the Technical Specification required testing of Emergency Diesel

Generator (EDG) fuel oil using more recent editions of the American

Society for Testing and Materials (ASTM) standards listed in

Technical Specification Surveillance Requirements 4.8.1.1.2e. and

4.8.1.1.2f. Prior to changing the edition of the previously approved

ASTM standard being used to evaluate the EDG fuel oil, the proposed

edition standard will be evaluated pursuant to 10 CFR Sec. 50.59,

``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits

a licensee to make changes in the procedures as described in the

safety analysis report without prior Commission approval, provided

that the proposed changes [do] not involve an unreviewed safety

question. Title 10 CFR Sec. 50.59(a)(2) states that a proposed

change involves an unreviewed safety question (iii) if the margin of

safety as defined in the basis for any technical specification is

reduced. Consequently, since any change to the edition of the ASTM

standard to be used to evaluate EDG fuel oil acceptability must be

evaluated relative to the more restrictive evaluation criterion of

10 CFR Sec. 50.59, then operation of the facility in accordance with

the proposed amendments would not involve a significant reduction in

a margin of safety.

ACTION statement g. of TS 3.8.1.1 is added to address the

required action in the event the new fuel oil properties do not meet

the Diesel Fuel Oil Testing Program limits. A failure to meet the

API gravity, kinematic viscosity, flash point or clarity limits is

cause for rejecting the new fuel oil prior to the addition to the

Diesel Fuel Oil Storage Tanks, but does not represent a failure to

meet the Limiting Condition for Operation (LCO) of TS 3.8.1.1, since

the new fuel oil has not been added to the storage tanks. Provided

these new fuel oil properties are met subsequent to the addition of

the new fuel oil to the storage tanks, 30 days is provided to

complete the analyses of the other fuel oil properties specified in

Table 1 of ASTM-D975-81, except sulfur which may be performed in

accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the

other new fuel oil properties specified in Table 1 of ASTM-D975-81

are not met, ACTION statement g. of TS 3.8.1.1 provides an

additional 30 days to meet the Diesel Fuel Oil Testing Program

limits. This additional 30 day period is acceptable because the fuel

oil properties of interest, even if they are not within limits,

would not have an immediate effect on EDG operation.

ACTION statement h. of TS 3.8.1.1 is added to address the

required action in the event the stored fuel oil total particulates

[do] not meet the Diesel Fuel Oil Testing Program limits. Fuel oil

degradation during long term storage shows up as an increase in

particulate, due mostly to oxidation. The presence of particulate

does not mean the fuel oil will not burn properly in a diesel

engine. The frequency for performing surveillance on stored fuel oil

is based on stored fuel oil degradation trends which indicate that

particulate concentration is unlikely to change significantly

between surveillances.

Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil

Testing Program, FPL will need to determine if the proposed program

change is at least as, if not more, effective, in detecting

unsatisfactory fuel oil. Since the proposed changes will require a

safety evaluation to assure that the reliability of the EDGs using

fuel oil tested in accordance with the different ASTM standard

edition maintains the current margin of safety, the proposed changes

do not involve a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,

P.C., 1615 L Street, NW., Washington, DC 20036

NRC Project Director: Victor M. McCree, Acting

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,

Burke County, Georgia

Date of amendment request: April 28, 1994

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3/4.8.1.1, ``AC Sources

Operating,'' and the associated TS Bases for demonstrating the

operability of the diesel generators (DGs), based upon three NRC

guidelines:

A. Generic Letter (GL) 93-05, ``Line-Item Technical Specifications

Improvements to Reduce Surveillance Requirements for Testing During

Power Operation.''

1. Delete from TS action statement a the requirement to test the

DGs in the event of an inoperable offsite circuit.

2. Eliminate from TS action statement b the need to test the

operable DG if the other DG became inoperable due to an inoperable

support system or an independently testable component in addition to

the existing provision excluding preplanned preventive maintenance or

testing. Furthermore, ifthe operable DG must be tested, it would be

tested within 8 hours (rather than 24 hours) unless the absence of any

potential common mode failure for the remaining DG is demonstrated.

3. Also eliminate from TS action statement c the need to test the

operable DG if the other DG became inoperable due to an inoperable

support system or an independently testable component in addition to

the existing provision excluding preplanned preventive maintenance or

testing. In addition, the operable DG would not have to be tested if

the absence of any potential common mode failure for the remaining DG

is demonstrated. A reference to TS action statement a would be deleted

because of the proposed change to TS action statement a described

above.

4. Eliminate from TS action statement e the need to test the DGs

when two offsite circuits are inoperable.

5. Revise TS 4.8.1.1.2.g.2 to allow the DG to be gradually loaded,

as opposed to a fast loading of 60 seconds or less to an indicated

value of 6100-7000 kw. This change would extend gradual loading of DGs

(that GL 93-05 recommends for routine monthly surveillance) to the 6-

month surveillance.

B. Regulatory Guide (RG) 1.9, Revision 3, ``Selection, Design,

Qualification, and Testing of Emergency Diesel Generator Units Used as

Class 1E Onsite Electric Power Systems at Nuclear Power Plants,''

(insofar as this guide relates to reducing DG stress and wear due to

testing and the elimination of certain reporting requirements).

1. Incorporate into TS 4.8.1.1.2.a.4 the provision to perform

routine monthly testing by gradually accelerating the DG to operating

speed, rather than requiring the DG to attain rated voltage and

frequency within 11.4 seconds. As a direct result of this proposed

change, TS action statements b, c, and f would be revised to reference

TS 4.8.1.1.2.g.1 instead of TS 4.8.1.1.2.a.4 in the event that an

operable DG must be tested when the other DG is inoperable. This has

the effect of requiring the operable DG to be fast-started for testing

pursuant to the action statement.

2. In TS 4.8.1.1.2.h.7, separate the 24-hour endurance run from the

hot restart test. As a result, create new TS 4.8.1.1.2.h.8 to require

the hot restart test. The DG would be operated for a minimum of 2 hours

at a load of 6800-7000 kw, and the DG would be shut down. Within 5

minutes of shutdown, the DG would be restarted and required to attain

rated voltage and frequency within 11.4 seconds. Delete existing TS

footnote (which provides for reperforming the hot

restart test without repeating the 24-hour endurance test) which is no

longer required. Renumber existing TSs 4.8.1.1.2.h.8, .9, .10, .11, and

.12 to accommodate the addition of new TS 4.8.1.1.2.h.8.

3. Delete TS 4.8.1.1.3, ``Reports.'' (This is also in accordance

with the Improved Technical Specifications, Revision 0, dated September

28, 1992).

4. In TS 4.8.1.2, ``A. C. Sources Shutdown,'' delete the reference

to deleted TS 4.8.1.1.3.

C. NUMARC 87-00, Revision 1, ``Guidelines and Technical Bases for

NUMARC Initiatives Addressing Station Blackout at Light Water

Reactors,'' (insofar as it relates to the test frequency for a problem

DG). Specifically, TS Table 4.8-1, ``Diesel Generator Test Schedule,''

would be revised to incorporate the test schedule of Section D.2.4.4 of

Appendix D to NUMARC 87-00, Revision 1. Under the proposed schedule,

testing pursuant to TS 4.8.1.1.2.a would be conducted monthly provided

the number of valid failures in the last 25 demands for a given DG is

no more than 3. If the number of valid failures is 4 or more, testing

would be conducted at least once per 7 days (but at intervals of no

less than 24 hours) until 7 consecutive failure-free starts from

standby conditions and load-run demands have been performed. Note that

both NUMARC 87-00, Revision 1, and RG 1.9, Revision 3, use and define

the terms ``start demand, start failure, load-run demand, and load-run

failure'' rather than the old RG 1.108 terminology of valid tests. In

fact, Section D.2.4.4 of Appendix D to NUMARC 87-00 refers to the last

25 ``demands'' rather than tests. Therefore, the proposed change to TS

Table 4.8-1 would count valid failures in terms of demands rather than

valid tests. The criteria for determining the number of valid failures

and demands would be in accordance with RG 1.9, Revision 3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes affect the required actions in response to

inoperable offsite and onsite ac sources, surveillance requirements

for the emergency diesel generators, and reporting requirements for

diesel generator failures. The proposed changes are based on the

recommendations of Regulatory Guide 1.9, Revision 3, NUMARC 87-00,

Revision 1, and Generic Letter 93-05. They are expected to result in

improvements in diesel generator testing and failure reporting and

reduce diesel generator aging due to excessive testing. As such, the

proposed changes should result in improved diesel generator

reliability, thereby providing additional assurance that the diesel

generators will be capable of performing their safety function.

Therefore, the proposed changes will not significantly increase the

probability or consequences of any accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated. The proposed changes affect the action and surveillance

requirements for the onsite and offsite ac sources. Accordingly, the

proposed changes do not involve any change to the configuration or

method of operation of any plant equipment, and no new failure modes

have been defined for any plant system or component nor has any new

limiting failure been identified as a result of the proposed

changes. Therefore, the proposed changes do not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety. The proposed changes are based on existing

regulatory guidance. Under the proposed changes, the emergency

diesel generators will remain capable of performing their safety

function, and the effects of aging on the diesel generators will be

reduced by eliminating unnecessary testing. 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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

Attorney for licensee: Mr. Arthur H. Domby, Esquire, Troutman

Sanders, Nations Bank Plaza, Suite 5200, 600 Peachtree Street, NE.,

Atlanta, Georgia 30308-2210.

NRC Project Director: Herbert N. Berkow

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island

Nuclear Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania

Date of amendment request: October 9, 1991.

Description of amendment request: Facility Operating License No.

DPR-73, a possession only license for the TMI-2 facility, held by

General Public Utilities Nuclear Corporation (GPU Nuclear), expires

November 4, 2009. The proposed amendment would extend the expiration

date of

Facility Operating License No. DPR-73 for TMI-2 to April 19, 2014.

No other changes to the license or the Technical Specifications are

proposed.

The TMI-2 facility is currently in long term storage. GPU Nuclear,

the licensee, has named this storage period Post Defueling Monitored

Storage or PDMS. The licensee plans to maintain TMI-2 in PDMS until

Three Mile Island Nuclear Station Unit No. 1 (TMI-1) permanently ceases

operation, at which time both TMI-1 and TMI-2 will be decommissioned

simultaneously. The TMI-1 Operating License expires on April 19, 2014.

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:

10 CFR 50.92 provides the criteria which the Commission uses to

perform a No Significant Hazards Consideration. 10 CFR 50.92 states

that an amendment to a facility license involves No Significant

Hazards if operation of the facility in accordance with the proposed

amendment would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, or

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

from any accident previously evaluated, or

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

The proposed modification of the expiration date of the TMI-2

License does not involve any physical changes to the facility. All

that is involved is an extension of the time TMI-2 would be in a

monitored storage condition. Based on this, GPU Nuclear concludes

that the proposed change does not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. Accident

evaluations for PDMS are provided in the PDMS Safety Analysis Report

(SAR) and the PDMS Final Programmatic Environmental Impact

Statement, Supplement 3 (PEIS) dated August 1989. These documents

evaluated monitored storage of TMI-2 for extended periods of time

and provide for surveillances to ensure monitored storage conditions

are appropriately maintained. The PDMS PEIS specifically evaluated

monitored storage until 2014. No evaluated accident has a

probability or consequence that are increased significantly during

the period 2009 to 2014 over the period before 2009. Therefore, it

can be concluded that this change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

from any accident previously evaluated. As previously stated, the

act of modifying the expiration date of the TMI-2 License does not

involve any physical changes to the facility and therefore, the

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

3. Involve a significant reduction in the margin of safety

during PDMS. The surveillances identified in the PDMS SAR will be

performed to ensure that the facility is maintained in the condition

defined by the SAR. These conditions and surveillances will continue

to apply during the extended license. Therefore, there will not be a

reduction in the margin of safety.

Based on the above analysis, it is concluded that the proposed

changes involve No Significant Hazards Consideration as defined by

10 CFR 50.92.

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: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601 Harrisburg, Pennsylvania 17105

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

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

NRC Project Director: Seymour H. Weiss

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,

Linn County, Iowa

Date of amendment request: June 30, 1994

Description of amendment request: The proposed amendment would

clarify the requirement for the audit of conformance to Technical

Specifications, delete the requirement for Safety Committee oversight

of the Emergency Plan and Security Plan and allow designation by the

Plant Superintendent signature authority for procedure approval.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

)The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. No physical changes will result from this amendment.

These changes revise audit requirements and procedure approval

requirements. The subject audits will still be performed to provide

assurance of conformance to the requirements, and the procedures

will still receive adequate technical reviews by the cognizant

departments while relieving the Plant Superintendent-Nuclear of the

administrative burden of signing each procedure revision.

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

new or different kind of accident from any accident previously

evaluated. No physical changes will result from this amendment.

These changes revise audit requirements and procedure approval

requirements. The subject audits will still be performed to provide

assurance of conformance to the requirements, and the procedures

will still receive adequate technical reviews by the cognizant

departments while relieving the Plant Superintendent-Nuclear of the

administrative burden of signing each procedure revision.

3) The proposed amendment does not involve a significant

reduction in a margin of safety. No physical changes will result

from this amendment. These changes revise audit requirements and

procedure approval requirements. The subject audits will still be

performed to provide assurance of conformance to the requirements,

and the procedures will still receive adequate technical reviews by

the cognizant departments while relieving the Plant Superintendent-

Nuclear of the administrative burden of signing each procedure

revision.

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

thisreview, 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: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401

Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,

Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC

20036

NRC Project Director: John N. Hannon

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,

Linn County, Iowa

Date of amendment request: June 30, 1994

Description of amendment request: The proposed amendment would add

Operability Requirements, Limiting Conditions for Operations (LCO) and

Surveillance Requirements for the Control Building Chillers.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated because the requested revisions do not affect

any FSAR analysis involving these systems.

The proposed revision only adds LIMITING CONDITIONS for

OPERATION (LCO) and Surveillance Requirements (SR) for the Control

Building Chillers. These additions will provide assurance that the

affected systems will be OPERABLE when required.

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

new or different kind of accident from any accident previously

evaluated because there is no equipment or design change associated

with this change. The proposed amendment only adds LCOs and SRs for

the Control Building Chillers.

3) The proposed amendment will not involve any reduction in a

margin of safety. The safety function of the Control Building

Chillers is to remove the design basis heat load under all normal

and emergency conditions. The addition of LCOs and SRs for the

Control Building Chillers ensures they will be OPERABLE when

required.

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

thisreview, 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: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401

Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,

Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC

20036

NRC Project Director: John N. Hannon

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,

Linn County, Iowa

Date of amendment request: June 30, 1994

Description of amendment request: The proposed amendment would

modify the surveillance testing of the Emergency Service Water (ESW)

system by deleting the flow rate test and the requirement to test the

pumps each week when river water temperature exceeds 80 deg.F and by

adding a surveillance regarding the Cedar River (Ultimate Heat Sink)

water temperature.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. No physical changes will result from this

amendment. The ESW system will still maintain its ability to support

various safety related equipment which is designed to mitigate the

consequences of certain accidents and transients. These safety

related systems play no part in the probability of these accidents

or transients occurring. Since the ESW system will continue to fully

support the cooling requirements of the safety related equipment

which mitigate the consequences of certain accidents and transients,

this amendment will not affect the consequences of these accidents

and transients. The re-analysis of the component heat loads assumed

worst case conditions and involved conservative assumptions. Our

continuing program for monitoring heat exchanger performance, which

was established in response to Generic Letter 89-13, ``Service Water

System Problems Affecting Safety-Related Equipment,'' will continue

to verify that the individual components are capable of performing

their design function. Therefore, the proposed amendment does not

involve a change in the probability or consequences of an accident

previously evaluated.

2) The proposed license amendment does not create the

possibility of a new or different kind of accident from any

previously evaluated. The safety function of the ESW system is

unchanged. The revised flow requirements for the system have been

established using conservative assumptions and worst case heat loads

and are appropriately documented in the FSAR and plant procedures.

This amendment will result in no physical changes to the ESW system

and therefore, will not affect its ability to continue to provide

reliable cooling water. Consequently, the proposed license amendment

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

accident from any previously evaluated.

3) The proposed amendment will not reduce the margin of safety.

The re-analysis of the ESW flow rate requirements and component heat

loads was performed using conservative assumptions and maximum

component heat loads. The actual operation of the ESW system will

not be changed. Any degradation of ESW pump performance would be

detected by the IST program which requires quarterly testing of

these pumps and monitoring of the pump's differential pressure and

flow. Deleting the requirement to perform the surveillance each week

when river water temperature exceeds 80 deg.F will not reduce the

margin of safety because even at a river water temperature of

95 deg.F, the required ESW flow to supply all the branches is well

below the normal system flow rate of approximately 1100 gpm.

Deleting the weekly surveillance will eliminate unnecessary testing

of the ESW pumps, thereby reducing wear on the pumps. Adding a

surveillance requirement for river water temperature will formalize

the recording of water temperature every hour to assure acceptable

ESW performance.

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

thisreview, 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: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401

Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,

Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC

20036

NRC Project Director: John N. Hannon

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

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

Date of amendment request: June 30, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.2.7.1, ``Primary Coolant System

Pressure Isolation Valves.'' Specifically, TS Table 3.2.7.1, ``Primary

Coolant System Pressure Isolation Valves,'' would be revised to add

Shutdown Cooling System (SCS) check valves 38-165, 166, 167, 168, 169,

170, 171, and 172 each with a maximum allowable leakage rate of less

than or equal to 0.375 gpm. The proposed amendment would add the check

valves in lieu of replacing the SCS isolation valves with ones that are

10 CFR Part 50, Appendix J, Type C air testable. The added check valves

would provide high pressure/low pressure interfaces between the high

pressure Reactor Coolant System and the low pressure Core Spray System.

The addition of the check valves will allow utilization of the Core

Spray System as a seal water system for sealing the Shutdown Cooling

isolation valves as permitted by Section III.C.3 of 10 CFR Part 50,

Appendix J.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change requires the addition of Primary Coolant

System pressure isolation valves for the prevention of an

intersystem LOCA [loss-of-coolant accident]. The proposed addition

does not affect operation of either the Shutdown Cooling or Core

Spray Systems. These changes do not alter any accident initiators or

precursors and therefore does not affect the probability of a

previously evaluated accident.

Testing these valves in accordance with Specification 3.2.7.1

provides assurance that the Core Spray System will not be damaged by

an overpressurization event which could lead to potential loss of

integrity of the system and subsequent release of radioactivity.

Thus, the addition of the valves would not increase the consequences

of any accident. Therefore, the operation of Nine Mile Point Unit 1,

in accordance with the proposed amendment, will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed addition of Primary Coolant System pressure

isolation valves, although a physical change, does not alter the

initial conditions used for any design basis accident. The check

valves provide the high pressure/low pressure isolation between the

Reactor Coolant and Core Spray Systems. These valves will be subject

to leak rate testing in accordance with Specification 3.2.7.1. This

ensures that an intersystem LOCA is prevented. The proposed change

has no effect on operation of either the Shutdown Cooling or Core

Spray Systems. Therefore, the design capabilities of these systems

are not challenged in a manner previously assessed so as to create

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

operation of Nine Mile Point Unit 1, in accordance with the proposed

amendment, will not create the possibility of a new or different

kind of accident from any accident previously analyzed.

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The proposed change which requires the addition of Primary

Coolant System pressure isolation valves, ensures proper isolation

of a high pressure/low pressure interface between the Reactor

Coolant and Core Spray Systems. The pressure isolation valves will

be leak tested in accordance with Specification 3.2.7.1. This

provides assurance that the Core Spray System will not be damaged by

an overpressurization event and will not result in loss of integrity

of the system. Thus, the results of any event previously analyzed

remains unchanged. Therefore, the operation of Nine Mile Point Unit

1, in accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &

Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Project Director: Michael L. Boyle

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

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

Connecticut

Date of amendment request: April 18, 1994

Description of amendment request: The proposed change will revise

the current surveillance frequency which verifies area temperature

limits at least once per 12 hours. The revised surveillance requirement

will verify area temperature limits at least once per 7 days when the

data-logger alarm is operable, or at least once per 12 hours when the

data-logger alarm is inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

...The proposed change does not involve an SHC [significant

hazards consideration] because the change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change reduces the frequency at which area

temperature monitoring must be verified when the temperature data-

logger alarm function is operable. For conditions where the

temperature data-logger alarm function is inoperable, the frequency

at which normal ambient temperature is verified remains unchanged.

In addition, the proposed change does not affect any system,

equipment, or component credited in any previous accident

evaluation, any environmental qualification or post-accident

profiles. Therefore, the proposed change will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any previously evaluated.

The proposed change does not alter or affect the design,

function, failure mode, or operation of the plant. There is no

change to the way in which the plant is operated and, therefore, no

increase in the probability of plant operation with any area

temperature outside of its limits. Therefore, this change does not

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

those previously evaluated.

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

The proposed change does not challenge or affect the performance

of any of the protective boundaries, revise temperature limits in

the technical specifications, or perform any modifications that

would increase the likelihood of technical specification temperature

limits being exceeded. The proposed change requires the data-logger

alarm function to be operable in order to relax the surveillance

frequency. This alarm function provides continuous monitoring that

would detect temperature excursions prior to the current

surveillance which does not credit operability of the data-logger

alarm function. Also, the proposed change does not increase the

interval for which temperatures could exceed technical specification

limits. Therefore, the proposed change does not cause a reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, City Place, Hartford, Connecticut 06103-3499.

NRC Project Director: John F. Stolz

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

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

Connecticut

Date of amendment request: June 30, 1994

Description of amendment request: The proposed change revises the

Technical Specifications to change the trip setpoint for the 4kV bus

undervoltage relay (for the grid degraded voltage) from its current

value of [greater than or equal to] 3710 volts to its new setting of

[greater than or equal to] 3730 volts.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

...The proposed change does not involve an SHC [significant

hazards consideration] because the change does not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change involves the modification of the

undervoltage relay setpoint from 3710V to 3730V. The protection

provided by this system in unaffected and is still in accordance

with the guidance provided in NRC Branch Technical Position PSB-1.

This refinement increases the Technical Specification minimum trip

setpoint for the degraded voltage relays on the 4kV safety buses. It

does not detrimentally affect the safe operation of the plant, nor

does this proposed modification increase the probability or

consequences of an accident previously evaluated. The actual trip

setpoints of the subject relays do not require any changes and are

currently conservatively set at 3745V. The allowable value of

[greater than or equal to] 3706V remains unchanged. This slightly

higher than required setting was chosen by NNECO to provide added

margin should an undervoltage condition be present. This higher

setting will not cause more actuations of the ESF [engineered safety

feature] systems.

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

from any accident previously evaluated.

The undervoltage protection system is provided to address the

concerns identified in NRC Branch Technical Position PSB-1 by

providing a scheme to detect the loss of offsite power at the class

1E buses, and a second level of undervoltage protection to protect

class 1E equipment. The change in the setpoint will not affect the

ability of this circuitry to detect a loss of offsite power or to

respond to an undervoltage condition.

Since the equipment will operate as previously described in the

FSAR [Final Safety Analysis Report], and there are no physical plant

modifications required (the current setting at the undervoltage

relay is 3745V), the proposed amendment 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.

These relays do not cause a loss of offsite power, nor do they

cause a degraded voltage condition. These relays react to conditions

that have been placed upon the Plant. In the event that a degraded

voltage condition exists on the 4kV safety buses, alarms in the

control room alert the operators of this condition. In addition, the

Connecticut Valley Electric Exchange (CONVEX), the system dispatch

center for generation and VAR/voltage control, is aware of the

minimum voltage requirements for the three nuclear plants at the

Millstone station and has a minimum target switchyard voltage of

345kV. Under normal operation conditions the switchyard voltage

would have to degrade below 328kV before one of the 4kV safety buses

would start to enter the degraded voltage level and trip the

degraded voltage circuit. These administrative controls help

preclude a degraded voltage condition on the 4kV safety buses prior

to actuation of the degraded voltage protection circuits.

The proposed change of the 4kV degraded voltage minimum trip

setpoint to 3730V from 3710V will not result in any physical relay

setting change. The existing trip setting for the 4kV degraded

voltage relays have been conservatively set at 3745V, while the

existing allowable value remains unchanged at 3706V.

The response times or actuation logic of the degraded voltage

protection circuit remains unaffected, therefore revising the trip

setpoint value in the Technical Specifications will not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, Connecticut 06360.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, City Place, Hartford, Connecticut 06103-3499.

NRC Project Director: John F. Stolz

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

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

County, Minnesota

Date of amendment requests: January 29, 1993, as revised June 15,

1994.

Description of amendment requests: The proposed amendments would

change core exit thermocouple action statements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The purpose of the post accident monitoring equipment is to

display unit variables that provide information required by the

control room operators during accident situations and as such help

limit the consequences of an accident. The proposed changes, which

will allow continued plant operation with less than four core exit

thermocouples per core quadrant, have no impact on the probability

of an accident because they are only used in response to accident

situations.

Continued plant operation with the core exit thermocouple system

in the degraded condition as allowed by the proposed core exit

thermocouple action statements would not affect the operators

ability to monitor for inadequate core cooling following an

accident. At least two core exit thermocouples would be operable per

core quadrant, a minimum of four thermocouples would be available in

the center region of the core and at least one thermocouple would be

available in each quadrant of the outside core region. The smaller

size of the Prairie Island core, and therefore higher density of

thermocouples per unit of core area, provides additional assurance

that core exit temperatures can be adequately monitored with a

reduced number of core exit thermocouples.

Alternate means of monitoring for inadequate core cooling would

also be available. These include the reactor vessel water level

indication system, the subcooling margin monitors and wide range

reactor coolant system temperature.

The combination of the remaining operable core exit

thermocouples and the alternate monitoring capability will ensure

that the operators ability to identify inadequate core cooling in a

timely manner and take appropriate corrective action will not be

impaired, and therefore; the proposed changes will have no

significant impact on the consequences of an accident.

The core exit thermocouples perform no active role in the

mitigation of an accident. Their inoperability will not affect the

operability of any engineered safety features equipment or that

equipments ability to mitigate the consequences of an accident.

Therefore, for the reasons discussed above, the proposed changes

will not significantly affect the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

analyzed.

There are no new failure modes or mechanisms associated with the

proposed changes. The proposed changes do not involve any

modification of plant equipment or any changes in operational

limits. The proposed changes only modify the requirements for

instrumentation used to monitor plant parameters during an accident.

The core exit thermocouples are passive monitoring devices, their

failure or inoperability cannot result in a plant accident of any

kind.

Therefore, for the reasons discussed above, the proposed changes

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

from any previously evaluated, and the accident analyses presented

in the Updated Safety Analysis Report will remain bounding.

3. The proposed amendment will not involve a significant

reduction in the margin of safety.

Continued plant operation with the core exit thermocouple system

in the degraded condition as allowed by the proposed core exit

thermocouple action statements would not affect the operators

ability to monitor for inadequate core cooling following an

accident. At least two core exit thermocouples would be operable per

core quadrant, a minimum of four thermocouples would be available in

the center region of the core and at least one thermocouple would be

available in each quadrant of the outside core region. The smaller

size of the Prairie Island core, and therefore higher density of

thermocouples per unit of core area, provides additional assurance

that core exit temperatures can be adequately monitored with a

reduced number of core exit thermocouples.

Alternative means of monitoring for inadequate core cooling

would also be available. These include the reactor vessel water

level indication system, the subcooling margin monitors and wide

range reactor coolant system temperature.

The combination of the remaining operable core exit

thermocouples and the alternate monitoring capability will ensure

that the operators ability to identify inadequate core cooling in a

timely manner and take appropriate corrective action will no be

impaired.

Therefore, for the reasons discussed above, the proposed changes

will not result in any reduction in the plant's 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 requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and

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

NRC Project Director: L. B. Marsh

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: June 30, 1994

Description of amendment request: This amendment would relocate

selected recirculation and control rod block instrumentation setpoints

from Technical Specifications (TS) Table 3.3.6-2, and Section 3/4.4.1

to the Core Operating Limits Report (COLR), thereby revising TS Section

6.9.1.9 to document relocation of these items into the COLR.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The TS change proposed is the relocation of the recirculation

pump Motor-Generator (MG) set mechanical and electrical stop and

control rod block recirculation flow upscale trip setpoint values to

the COLR. No physical plant equipment change is proposed. The TS

requirements for the setpoints and the associated surveillance

requirements remain unchanged. Only the location of the setpoint

values will be changed. The subject setpoint values will become

cycle depend[e]nt and will be determined by NRC approved methods, as

are the balance of setpoints and thermal limits found in the COLR.

However, the subject setpoint values are not modified as part of

this TS change.

Therefore, the proposed TS change does not involve an increase

in the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The TS changes proposed are the relocation of the recirculation

pump MG set mechanical and electrical stop and control rod block

recirculation flow upscale trip setpoint values to the COLR. No

physical plant equipment change is part of the proposed TS changes.

The TS LCOs and surveillance requirements remain unchanged. The only

change proposed is the relocation of the subject setpoint values as

noted above. These setpoint values have been determined in

accordance with previously NRC approved methods and assure

sufficient operating margins in accordance with existing core design

methodology. Therefore, the proposed TS changes do not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The following TS BASES were reviewed for potential reduction in

the margin of safety:

3/4.2 Power Distribution Limits

3/4.3.6 Control Rod Block Instrumentation

3/4.4.1 Recirculation System

The margin of safety, as defined in the TS BASES, will not be

reduced. The proposed TS changes do not affect existing accident

analyses or design assumptions, nor do they impact any safety limits

of the plant, since they are administrative in nature.

Therefore, the proposed TS changes do not involve a reduction in

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

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

Pennsylvania 19101

NRC Project Director: Charles L. Miller

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: June 23, 1994

Description of amendment request: This amendment will change the

Technical Specification 4.0.5 for each unit to reflect NRC's policy

with respect to relief requests for the inservice inspection programs.

Specifically, the change would clarify the fact that relief requests

for impracticable testing or surveillance requirements can be

implemented prior to the Commission approval of such requests.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes are administrative in nature in that the

changes eliminate any possibility of misinterpretation of the ASME

Code requirements that allow for a utility to submit relief requests

to the Commission within one year and allows for the implementation

of these request[s] prior to Commission review and approval. The

relief requests are based on and provide for alternative testing

based on industry practice that provides an equivalent level of

quality and safety as the Code requirement. The Commission will

still provide acceptance of the relief requests in writing.

Therefore, it can be concluded that the proposed changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

from any accident previously evaluated.

No new failure modes have been 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 changes do not create the possibility of

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

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

The proposed changes are administrative in nature and do not

adversely impact the plant's ability to meet applicable regulatory

requirements related to inservice testing or inspection. The

proposed changes eliminate any possible misinterpretation of the

Code requirements regarding relief requests and do not reduce the

protection of public health and safety. Therefore, it can be

concluded that 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: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

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

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

NRC Project Director: Charles L. Miller

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

50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: June 13, 1994

Description of amendment request: The proposed amendments would

permit an out-of-service component to be returned to service under

administrative controls for the purpose of determining operability. The

proposed change is consistent with the method utilized in the new

Westinghouse Standard Technical Specifications (NUREG-1431). In

addition, the proposed amendment corrects a typographical error in the

header information on Page 3/4 0-2 of the current technical

specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

consequences of an accident previously evaluated.

a. Header Information

This editorial change corrects a typographical error only. As

such, existing accident analyses are unaffected.

b. Specification 3.0.6

The proposed change merely clarifies the intent of Specification

3.0.2. As such, existing accident analyses are unaffected.

2. Will not create the possibility of a new or different kind of

accident from any accident previously evaluated.

a.Header Information

This editorial change corrects a typographical error only. As

such, it does not alter the function of any plant equipment, involve

any design changes, nor does it create any new operating modes or

accident scenarios.

b. Specification 3.0.6

The proposed change merely clarifies the intent of Specification

3.0.2. As such, it does not alter the function of any plant

equipment, involve any design changes, nor does it create any new

operating modes or accident scenarios.

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

safety.

a. Header Information

This editorial change corrects a typographical error only. As

such, the present margins of safety are unaffected.

b. Specification 3.0.6

The proposed change merely clarifies the intent of Specification

3.0.2. As such, the present margins of safety are unaffected.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW., Washington, DC 20005-3502

NRC Project Director: Charles L. Miller

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

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: June 17, 1994 as supplemented July 13,

1994

Description of amendment request: The proposed amendment would

change the requirement to perform the Channel Functional Test of the

Power Operated Relief Valve (PORV) position indication from quarterly

to every 18 months and to exempt the PORV Block Valve position

indication from performance of the channel Function Test if the PORV

Block Valve is shut as required to isolate a PORV that cannot be

manually cycled.

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 not involve a significant increase in the probability or

consequences of an accident previously evaluated.

A change from quarterly (Q) to at least every 18 months (R) may

appear to be non-conservative, at first; however, by extending the

surveillance requirement to cycle the PORV during non-power

conditions, the change eliminates the potential risk and

consequences of having the valve sticking open at power or not fully

closing (leaking). Therefore, by extending the surveillance the

probability and consequences of any previously analyzed accident is

reduced, since the testing would now be conducted in a non-power

condition, and the margin to safety is increased. Consequently, a

net safety gain is realized by eliminating or minimizing these

risks.

The added note for PORV block valve is included for consistency

and alignment between the surveillance requirement under this T/S

(Table 4.3-11) with that of T/S surveillances 4.4.3.2 and 4.4.5.2

for Units 1 and 2 respectively.

Therefore, the proposed amendment does not involve a physical or

procedural change to any structure, component, or system that

significantly affects accident/malfunction probabilities or

consequences previously evaluated in the UFSAR.

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

accident from any accident previously evaluated.

The proposed changes do not introduce any design or physical

configuration changes to the facility which could create new

accident scenarios.

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

safety.

As stated in response to question number 1 above, the proposed

changes do not eliminate the required T/S surveillance requirements.

The first change eliminates the need to cycle the PORV valves

through one complete cycle of full travel at power. The second

change allows for not having to perform a surveillance on a valve

that it is being used as an isolation point. The valve has been

closed to comply with requirements of another T/S action statement.

Therefore, the probability and consequences of any previously

analyzed accident is reduced, thus increasing the margin to safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW., Washington, DC 20005-3502

NRC Project Director: Charles L. Miller

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear

Plant, Unit 3, Limestone County, Alabama

Date of amendment request: March 29,1994 (TS 340)

Description of amendment request: The proposed amendment to the

Unit 3 Technical Specifications adds a limiting condition for operation

and a surveillance requirement for a load shedding logic being added by

a design change to Unit 3. The load shedding logic is being added to

ensure that the maximum capacity of the Unit 3 Emergency Diesel

Generators is not exceeded during a postulated loss of offsite power

event concurrent with a design basis accident.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

This proposed change establishes a surveillance testing

requirement and limiting condition for operation for the Unit 3 480-

volt load shedding logic system. This Technical Specification change

will not introduce any new failure mode and will not alter any

assumptions previously made in evaluating the consequences of an

accident. Accordingly, this change does not affect any design

limiting safety system settings or operating parameters.

Furthermore, the change does not modify or add any accident

initiating events or parameters. Therefore, these proposed changes

do not involve an increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

This proposed change establishes a limiting condition for

operation and a surveillance requirement for the Unit 3 480-volt

load shedding logic system. The addition of a limiting condition for

operation and surveillance requirement will not adversely affect the

operation of Unit 3 or the manner in which it is operated.

Furthermore, the change does not create a failure mode that can lead

to an accident of a different type than previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The addition of a limiting condition for operation and

surveillance requirement will not reduce the margin of safety. The

testing of the 480-volt load shedding logic on an 18-month interval

is consistent with BWR/4 (NUREG-1433) Standard Technical

Specifications. These are based on the guidance set forth in NRC

Regulatory Guide 1.108, ``Periodic Testing of Diesel Generator Units

Used as Onsite Electric Power Systems at Nuclear Power Plants.'' The

addition of a limiting condition for operation establishes a minimum

acceptable level of performance for the 480-volt load shedding logic

system. Thus, the ability of the Emergency Diesel Generators to

supply power during a loss of offsite power coincident with a design

basis accident is assured.

Furthermore, no reductions in the requirements or setpoints of

the equipment supplied by the Emergency Diesel Generators are made

which could result in a reduction in the margin of safety.

Therefore, this proposed change does not involve a reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: February 23, 1994

Brief description of amendments: The proposed changes would revise

the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2,

Technical Specifications (TS) to (1) allow a one-hour allowed outage

time (AOT) following discovery of a closed cold leg injection

accumulator discharge isolation in Modes 1, 2, or 3; (2) eliminate the

redundant requirement to reverify accumulator boron concentration

following fill from the refueling water storage tank (RWST); (3)

relocate the accumulator water level and pressure channel analog

channel operational test (ACOT) and channel calibration from the CPSES

Technical Specifications to an administratively controlled program; (4)

change the accumulator limits to analysis values rather than indicated

values; and (5) reduce the inspection frequency following containment

entries.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of a previously evaluated accident.

The current requirement to immediately open a cold leg

accumulator discharge isolation valve (or shut down the unit) upon

discovery that the valve is closed is modified by the requested

change to provide a one hour allowed outage time (AOT) prior to

requiring a unit shutdown. This change is consistent with NUREG-

1431. The currently required action is more restrictive than that

required by CPSES Technical Specification 3.0.3, that specifies the

action required if an LCO [Limiting Condition for Operation] and its

associated action requirements are not met and which provides a one

hour AOT prior to taking steps to place the plant in Mode 3 within

the following 6 hours. Following this requested change, the required

actions for an accumulator declared inoperable due to a closed

discharge isolation valve will be identical to those actions

required for inoperability for other reasons, with the exception of

the accumulator boron concentration being out of specification that

has an AOT of 72 hours. Changing the AOT from ``immediate'' to one

hour does not affect the probability of an accident. The only

previously evaluated accident that is potentially impacted is the

Loss of Coolant Accident (LOCA). With all valves open and thus all

accumulators available, a potential LOCA is bounded by the existing

accident analyses. With one accumulator discharge isolation valve

closed and thus one accumulator not available, the consequences of a

LOCA could be more severe; however, this requested amendment does

not create this scenario. In other words, although the change in AOT

may slightly increase the probability that, were a LOCA to occur, an

accumulator would not be available (see the response to [number] 3

below), it does not involve a significant increase in the

consequences of an accident previously evaluated.

The requirement to test the accumulator boron concentration

following a 101 gallon or greater solution volume increase is

modified by the requested change to exclude volume additions from

the Refueling Water Storage Tank (RWST). Since the RWST boron

concentration must be confirmed to satisfy the limits for the

accumulators, there is no impact on the probability or consequences

of any accident.

The relocation of the accumulator water level and pressure

channel ACOT and Channel Calibration from CPSES Technical

Specifications to an administratively controlled program is

essentially an administrative change. Because proper tests will

still be performed, there is no impact on the probability or

consequences of any accident.

The requested change to reduce the containment debris

inspections from ``at the completion of every entry'' to ``[a]t

least once daily'' will require fewer inspections and is consistent

with SR [Surveillance Requirement] 4.6.1.3 for the containment air

locks. The accident of concern is a LOCA and these inspections have

no impact on the probability of a LOCA. Performing fewer inspections

would slightly increase the possibility that, should a LOCA occur,

there could be debris in containment which could be transported to

and partially clog the containment sump. However, inspecting at

least daily if containment entries have been made is adequate and is

justified by the reduced total radiation exposure for plant

personnel. The inspections conducted at least daily assures that

there is not a significant increase in the consequence of any

accident.

The requested changes do not modify the existing LCOs for

Technical Specifications 3.5.1 and 3.5.2 with the exception of the

replacement of ``indicated'' values with analysis values in LCO

3.5.1, consistent with the relocation of the SRs for accumulator

instrumentation. The requested changes are consistent with NUREG-

1431 and GL 93-05, and, as such, have already been generically

assessed by the NRC. It is concluded that the requested changes do

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any previously evaluated

accident.

The only requested change that modifies current operation of the

plant is the requested one hour alowed outage time for action

following discovery of a closed cold leg accumulator discharge

isolation valve. The requested one hour completion time to open the

valve continues to ensure that prompt action will be taken to return

the inoperable accumulator to an operable status, minimizing the

potential for exposure of the plant to a LOCA under this condition.

In addition, as LCO 3.5.1a will continue to require that the

accumulator discharge isolation valve be open with power removed

from the valve operator, the probability of the discharge isolation

valve being closed in Modes 1, 2, or 3 will remain low. This change

in current operation does not create the possibility of a new or

different kind of accident.

The requested slight reduction in the containment inspection

frequencies specified in SR 4.5.2 only serves to reduce the number

of unnecessary inspections. It does not make substantial changes to

the inspection requirements, nor does it change the method of

performing these requirements. Thus, the requested change does not

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

No significant changes to the limiting conditions for operation

of the accumulators or the emergency core cooling system are

requested as part of this amendment request. The requested changes

do not involve any physical changes to the plant. The requested

changes are consistent with NUREG-1431 and GL 93-05, and, as such,

have already been generically assessed by the NRC. Thus, the

requested changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The only requested change that modifies current operation of the

plant is the requested one hour allowed outage time for action

following discovery of a closed cold leg accumulator discharge

isolation valve. As noted in the response to [number] 1 above, this

requested change in AOT does not significantly affect the

probability or consequences of an accident, but does increase the

possibility that, should a LOCA occur, one of the accumulators may

not be available to help mitigate the consequences of the accident.

However, the requested one hour completion time to open the valve

continues to ensure that prompt action will be taken to return the

inoperable accumulator to an operable status, minimizing the

potential for exposure of the plant to a LOCA under this condition.

In addition, as LCO 3.5.1a will continue to require that the

accumulator discharge isolation valve be open with power removed

from the valve operator, the probability of the discharge isolation

valve being closed in Modes 1, 2, or 3 will remain low. Considering

the controls above and the fact that the requested action statement

is consistent with TS 3.0.3, it is concluded that the requested

change does not involve a significant reduction in the margin of

safety.

The requested slight reduction in the containment inspection

frequencies specified in SR 4.5.2 only serves to reduce the number

of unnecessary inspections conducted and reduce the personnel

exposure associated with the inspections. As adequate inspections

will continue to be conducted, this requested change does not

involve a significant reduction in a margin of safety.

No significant changes to the limiting conditions for operation

of the accumulators or the emergency core cooling systems are

requested as part of this amendment request. The requested changes

are consistent with NUREG-1366, NUREG-1431 and GL 93-05, and, as

such, have already been generically assessed by the NRC. Thus, it is

concluded that the requested changes do not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.

Box 19497, Arlington, Texas 76019

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036

NRC Project Director: William D. Beckner

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

Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: April 22, 1994

Brief description of amendments: The proposed amendments would

revise the technical specifications by changing the frequency of

auxiliary feedwater pump operational testing from monthly to quarterly.

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 or a previously evaluated

accident.

Because the Auxiliary Feedwater System pumps are provided to

mitigate certain accidents, altering the test frequency of the pumps

will not impact the probability of an accident. The Auxiliary

Feedwater System pumps will continue to be tested quarterly on a

staggered basis to the same standards applied to safety-related

pumps as defined by ASME Section XI. Satisfactory completion of the

testing in accordance with the Code is used as verification that

safety-related pumps will be available to perform their intended

function. Quarterly testing of the Auxiliary Feedwater System pumps

on a staggered basis, therefore, will continue to assure that the

Auxiliary Feedwater System will be capable of performing its

intended function. It is thus concluded that the requested change

will not involve a significant increase in the probability of

occurrence or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

Changing the surveillance test frequency of the Auxiliary

Feedwater Pumps does not involve any physical modification of the

plant or result in a change in a method of operation. Therefore, it

is concluded that the requested change does not create the

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

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

Changing the surveillance testing frequency of the Auxiliary

Feedwater System pumps does not affect any safety limits or any

limiting safety system settings. System operating parameters are

unaffected. The availability of equipment required to mitigate or

assess the consequence of an accident is not reduced; in fact the

availability is increased because the system is rendered inoperable

on a quarterly basis to perform pump testing, rather than a monthly

basis. Further, vibration testing being the most effective early

indication of gradual pump degradation continues to be performed on

the same frequency. Quarterly testing of the Auxiliary Feedwater

pumps on a staggered basis in accordance with the criteria specified

in the ASME Section XI code provides adequate assurance that the

Auxiliary Feedwater System pumps are capable of performing their

intended function. Thus, its [sic] is concluded that the requested

change does not involved [sic] a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.

Box 19497, Arlington, Texas 76019

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036

NRC Project Director: William D. Beckner

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

Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: April 25, 1994

Brief description of amendments: The proposed amendments would

revise the technical specifications to reduce the number of fast starts

currently required by surveillanc

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