Biweekly Notice

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from September 19, 1994, through September 29,

1994. The last biweekly notice was published on September 28, 1994 (59

FR 49425).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the 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 November 14, 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, et al., Docket No. 50-400, Shearon

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

North Carolina

Date of amendment request: August 19, 1994

Description of amendment request: The proposed amendment will move

the current procedural details of the radiological effluent Technical

Specifications (TS) programmatic controls for radioactive effluents,

radiological environmental monitoring and solid radioactive wastes from

the Administrative Controls Section of the TS to the Offsite Dose

Calculation Manual (ODCM) or the Process Control Program (PCP), as

appropriate, in accordance with the guidance of Generic Letter 89-01.

This amendment will also incorporate changes to the reporting

requirements for the Effluent Release Reports, in accordance with 10

CFR 50.36; incorporate references to the new 10 CFR Part 20; and revise

the terminology for the gaseous effluent release rate limits.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Transferring the procedural details from the TS to the ODCM and

PCP and their replacement with programmatic controls have no impact

on plant operation or safety. No safety-related equipment, safety

function, or plant operation will be altered as a result of this

proposed change. The changes are unrelated to the initiation and

mitigation of accidents and equipment malfunctions addressed in the

Final Safety Analysis Report.

The proposed revisions to the reporting requirements for

Effluent Release Reports, the gaseous effluent release rate limit

and the relocation of the old 10 CFR 20.106 requirements to the new

10 CFR 20.1302 have no impact on plant systems, plant operations or

accident precursors. The changes to the Effluent Report requirements

and the updated reference to 10 CFR 20.1302 are administrative in

nature. The change to the gaseous effluent release limit is also

administrative in nature in that it will allow the continued

operation of the facility with the same release rate limits as are

currently implemented by the Technical Specifications.

Therefore, there would be no 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.

Transferring the procedural details from the TS to the ODCM and

PCP and their replacement with programmatic controls have no impact

on plant operation or safety. No safety-related equipment, safety

function, or plant operation will be altered as a result of this

proposed change. No changes to plant components or structures are

introduced which could create new accidents or malfunctions not

previously evaluated.

The proposed revisions to the reporting requirements for

effluent Release Reports, the gaseous effluent release rate limit

and the relocation of the old 10 CFR 20.106 requirements to the new

10 CFR 20.1302 have no impact on plant systems, plant operations or

accident precursors. The changes to the Effluent Report requirements

and the updated reference to 10 CFR 20.1302 are administrative in

nature. The change to the gaseous effluent release limits is also

administrative in nature in that it will allow the continued

operation of the facility with the same release rate limits as are

currently implemented by the Technical Specifications.

Therefore, the proposed changes do 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 the margin of safety.

The procedural details of the current RETS will be transferred

to the ODCM and PCP and replaced with programmatic controls

consistent with regulatory requirements, including controls on

revisions to the ODCM and PCP. Thus, no requirements or controls

will be reduced.

The changes to the Effluent Report requirements and the updated

reference to 10 CFR 20.1302 are administrative in nature and

therefore have no effect on the margin of safety. The proposed

revisions to the gaseous effluent release limits will maintain the

release rate limits at the same level as currently implemented by

the Technical Specifications. Therefore, there will be no change in

the types and amounts of effluents that will be released, nor will

there be an increase in individual or cumulative radiation exposures

to any member of the public.

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: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of amendment request: September 19, 1994

Description of amendment request: The proposed amendment would

revise the Technical Specifications by reducing the frequency for

testing the containment spray system spray nozzles.

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 of occurrence or consequences of any accident

previously evaluated.

The relaxation of surveillance frequency will not affect any of

the initiators or precursors of any accident previously evaluated.

Performance of CS spray nozzle testing on a ten year basis rather

than on a five year basis will not increase the likelihood that a

transient initiating event will occur because transients are

initiated by external events, equipment malfunction, and/or

catastrophic system failure. There are no failure mechanisms or

modes for the CS system or spray nozzles that could initiate a

transient since the CS system is passive except during a Loss of

Coolant Accident (LOCA). Upon receipt of a Containment Spray signal

(Containment High-High Pressure coincident with a Safety Injection

Signal), the CS pumps automatically start and valves align to

provide spray flow through the CS risers, ring headers, and out the

spray nozzles. Periodic testing requirements for the CS pumps and

valves (the active components of the system) are unaffected by the

proposed changes. Industry experience and previous test experience

at Zion Station supports the conclusion that functional checks of

the spray nozzles on a ten year basis is adequate to detect

degradation or blockage of the spray nozzles.

The proposed typographical and administrative changes do not

affect the operability or surveillance requirements given in

Technical Specifications. They will only improve consistency of

existing terminology and format of Technical Specifications and will

remove temporarily imposed Bases that are no longer applicable.

Based on the fact that reliability of the system will not be

affected and transient precursors and initiators are not affected by

operation in accordance with the proposed changes, the probability

of occurrence of accidents previously evaluated will not

significantly increase.

The proposed change in surveillance frequency will not affect

the ability of the CS system to function as designed during the

accidents considered in the Safety Analyses. Periodic testing

requirements for the CS pumps and valves (the active components of

the system) are unaffected by the proposed changes. Industry

experience and previous test experience at Zion Station supports the

conclusion that functional checks of the spray nozzles on a ten year

basis is adequate to detect degradation or blockage of the spray

nozzles. Given the proposed changes, the CS system will maintain the

ability to reduce containment pressure, remove heat from

containment, and remove iodine from the containment atmosphere

during the design basis LOCA. As a result, peak containment pressure

will be maintained below design pressure and the off-site release

due to the postulated accident will remain as described in the

Safety Analyses. Therefore, based on the previous discussion, the

proposed changes do not involve a significant increase in

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 previously analyzed.

The proposed changes to the Technical Specifications do not

involve the addition of any new or different types of safety related

equipment, nor does it involve the operation of equipment required

for safety operation of the facility in a manner different from

those addressed in the safety analyses. No safety related equipment

or function will be altered as a result of the proposed changes.

Also, changes to the procedures governing normal plant operation and

recovery from an accident are not necessitated by the proposed

Technical Specification changes.

The proposed typographical and administrative changes do not

affect the operability or surveillance requirements given in

Technical Specifications. They will only improve consistency of

existing terminology and format of Technical Specifications and will

remove Bases that are no longer applicable.

Since no new failure modes or mechanisms are added by the

proposed changes, the possibility or a new or different kind of

accident is not created.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

Plant safety margins are established through LCOs, limiting

safety system settings, and safety limits specified in the Technical

Specifications. There will be no changes to either the physical

design of the plant or to any of these settings and limits as a

result of relaxing the surveillance frequency of CS nozzle checks

from five years to ten years. Testing on a ten year basis is

adequate to detect spray nozzle degradation or blockage since the

system piping and spray nozzles are constructed of corrosion

resistant Type 304 stainless steel and since the system is normally

passive (i.e. spray risers and spray rings are empty with no flow

except during an accident). This conclusion was also provided in

NUREG-1366 and Generic Letter 93-05 which recommended revising the

surveillance frequency as proposed.

The proposed typographical and administrative changes do not

affect the operability or surveillance requirements given in

Technical Specifications. They will only improve consistency of

existing terminology and format of Technical Specifications and will

remove Bases that are no longer applicable.

Based on the above discussion, the ability to safely shut down

the operating unit and mitigate the consequences of all accidents

previously evaluated will be maintained. Therefore, the margin of

safety is not significantly affected.

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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085

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: August 25, 1994

Description of amendment request: The requested amendments modify

the trip setpoint and allowable value for the 4 kilo-volt (KV)

electrical bus degraded grid undervoltage relay and the allowable value

for the loss of offsite power relay in response to an issue identified

in the licensee's Self-Initiated Technical Audit.

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. The amendments will not affect either the probability or

the consequences of an accident, since no physical changes to the

plant are being proposed. The amendments merely change the existing

technical specification settings for the above relays to more

conservative values. Current field settings for these relays are

already at these more conservative values. No changes to the manner

in which the plant is operated are being proposed.

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, no actual changes to the physical plant

are being proposed. No effect on plant operation will occur,

therefore the possibility of new accident types is not created.

Criterion 3

The requested amendments will not involve a significant

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

unaffected, since no changes to the plant are being made. The

proposed technical specification values are more conservative and

are intended to make the technical specifications correspond with

the actual plant relay settings.

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

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

Date of amendment request: August 25, 1994

Description of amendment request: The amendments would change the

frequency for conducting the surveillance test required by TS 4.7.1.2.1

for the auxiliary feedwater pumps from once per 31 days to at least

once per 92 days and would add a footnote which clarifies that testing

is not required to be performed until system heatup has progressed to a

pressure (600 psig) that will support conduct of the test. The change

in the surveillance frequency has been evaluated and approved by the

NRC staff as discussed in Section 9.1 of NUREG-1366, ``Improvements to

Technical Specifications Surveillance Requirements.'' The change is

based on the finding in NUREG-1366 that an analysis of AFW pump

failures indicates that a monthly surveillance test interval may be

contributing to AFW pump unavailability through failures and equipment

degradation and, therefore, AFW pump availability is increased by

quarterly testing on a staggered basis. Generic Letter 93-05, ``Line-

Item Technical Specification Improvements to Reduce Surveillance

Requirements for Testing During Power Operation,'' provided the sample

TS for this change. The change is accomplished by dividing TS

4.7.1.2.1a into two parts. The new 4.7.1.2.1a maintains the previous

31-day testing frequency for the AFW valves while the new 4.7.1.2.1b

inserts a new frequency of once per 92 days for the AFW pump tests.

Also, an obsolete footnote is deleted. The new footnote discussed above

is consistent with NUREG-1431, ``Standard Technical Specifications for

Westinghouse Plants.'' Appropriate changes to the Bases for the TS have

also been proposed.

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 requested amendments decrease from monthly to quarterly the

frequency at which the motor-driven and turbine-driven AFW pumps

must be demonstrated operable as specified in TS 4.7.1.2.1. They

also incorporate a note of clarification from the new Westinghouse

STS into the existing Catawba specifications concerning when the

pump head or discharge pressure versus flow verification for the

turbine-driven pump is required to be performed.

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Decreasing the frequency of AFW pump testing as specified

in TS from monthly to quarterly will have no impact upon the

probability of any accident, since the AFW pumps are not accident

initiating equipment. Also, since Catawba's AFW pump performance

history supports 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 footnotes as indicated in the

Catawba technical specification markups is purely an administrative

change, and therefore will have no impact upon either the

probability or consequences of any accident. Incorporating the new

STS note will only serve to clarify when the turbine-driven pump is

required to be tested and will not have any impact upon either the

probability or consequences of any accident. The pump will still be

tested as before and its acceptance criteria will be unaffected.

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 AFW pumps are 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 footnotes has no bearing on

any accident initiating mechanisms. Incorporating the clarifying

note from the new STS will not result in any new acident sequences,

since plant operation will be unaffected.

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 AFW pumps will still be

capable of fulfilling their required safety function, since plant

operating experience supports the proposed change. The availability

of the AFW pumps will be increased as a result of the proposed

amendments because they will not have to be made unavailable for

testing as frequently. 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

footnotes will not result in any impact to plant safety margins.

Incorporating the note from the new STS will not impact any safety

margins.

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

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

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

Date of amendment request: September 1, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.2.2, ``Minimum Reactor Vessel

Temperature for Pressurization,'' and the associated Bases.

Specifically, the proposed amendment replaces existing TS Figures

3.2.2.a,b,c,d, and e and associated TS Tables 3.2.2.a,b,c,d, and e,

that define the limits for minimum reactor vessel temperature for

pressurization and account for neutron damage at exposures up to 18

effective full power years (EFPY), with new figures and tables that are

applicable for up to 18 EFPY. The licensee stated that the new

pressure-temperature (P-T) limits were developed based on a plant-

specific Charpy shift model for Nine Mile Point Nuclear Station Unit

No. 1 which is consistent with and meets the requirements of Regulatory

Guide 1.99, Revision 2, ``Radiation Embrittlement of Reactor Vessel

Materials.'' The new P-T limits were calculated in accordance with 10

CFR Part 50, Appendix G, and with the requirements specified in

Appendix G to Section III of the American Society of Mechnical

Engineers Boiler and Pressure Vessel Code (ASME Code).

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 [NMP1], in accordance

with the proposed amendment, will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Components of the reactor primary coolant system are operated so

that no substantial mechanical or thermal loading is applied unless

the reactor pressure vessel (RPV) materials are at a temperature

well above the reference nil-ductility temperature (RTNDT) of

the limiting RPV material. Protection against brittle fracture is

further ensured by postulating a defect with a depth 1/4 of the RPV

wall thickness and a length 1-1/2 times the wall thickness, and

calculating the allowable pressure loading as a function of

temperature using linear elastic fracture mechanics. Safety factors

are applied to the allowable loading determination and lower bound

fracture toughness properties are used to represent the material

behavior. The net effect of the 10 CFR [Part] 50, Appendix G and the

ASME Section III, Appendix G P-T curve calculative procedures is to

produce very conservative P-T curves. These procedures have been

applied in the calculation of the proposed P-T limits.

Neutron damage during plant operation is accounted for in the

allowable pressure loading by calculating an adjusted reference nil-

ductility temperature (ARTNDT). Regulatory Guide 1.99, Revision

2, defines the ARTNDT as the sum of the reference nil-ductility

temperature (RTNDT) plus the shift in the reference nil-

ductility temperature caused by irradiation ([delta]RTNDT),

plus a margin. The proposed amendment replaces Equation (2) in

Regulatory Position 2.1 with an accurate plant-specific model. The

ARTNDT margin is the same as for earlier P-T curve

calculations. Operation of NMP1 in accordance with the proposed P-T

operating limits will preclude brittle failure of the RPV materials.

Safety margins for brittle fracture are in accordance with those

specified in 10 CFR [Part] 50, Appendix G and Appendix G to Section

III of the ASME Code. Therefore, 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 amendment incorporates P-T operating limits based

on previously established calculative procedures described in 10 CFR

[Part] 50, Appendix G, Appendix G to Section III of the ASME Code,

and Regulatory Guide 1.99, Revision 2. The proposed changes to the

P-T operating limits are based on analyses of the irradiated

limiting plate material for Nine Mile Point Unit 1. The proposed

changes do not modify any plant equipment nor do they create any

potential initiating events that would create any new or different

kind of accident. Operation in accordance with the proposed P-T

operating limits will preclude brittle failure of the reactor vessel

material, since safety margins specified in 10 CFR [Part] 50,

Appendix G and Appendix G to Section III of the ASME Code will be

maintained. Therefore, the proposed P-T limits will not create the

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

previously evaluated.

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

amendment, will not involve a significant reduction in a margin of

safety.

Operation in accordance with the proposed P-T operating limits

will preclude brittle failure of the reactor pressure vessel since

safety margins in 10 CFR [Part] 50, Appendix G and Appendix G to

Section III of the ASME Code will be maintained. The plant-specific

limiting material [delta]RTNDT has been reduced as compared

with the overly conservative [delta]RTNDT used in previous P-T

curve calculations as a result of the more accurate representation

of the Nine Mile Point Unit 1 RPV plate behavior as a function of

neutron exposure. However, the [delta]RTNDT is intended to be

an accurate representation of the Charpy shift (indexed at 30 ft-lbs

of absorbed energy) as a function of fluence. Since the ASME Section

III, Appendix G safety factors have been maintained and the

Regulatory Guide 1.99, Revision 2, margin term specified in

Regulatory Position 2.1 has been applied in the same manner as in

earlier P-T curve calculations, no significant reduction in the

margin of safety has resulted from the use of a plant-specific

[delta]RTNDT model. Therefore, 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 J. Case

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

323,Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County, California

Date of amendment request: August 17, 1994 (Reference LAR 94-06)

Description of amendment request: The proposed amendments would

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

Power Plant (DCPP) Unit Nos. 1 and 2 to change TS 3/4.1.2.5, ``Borated

Water Sources - Shutdown,'' TS 3/4.1.2.6, ``Borated Water Sources -

Operating,'' and TS 3/4.5.5, ``Emergency Core Cooling Systems -

Refueling Water Storage Tank.'' The changes delete the minimum

refueling water storage tank (RWST) solution temperature and increase

the allowed outage time (AOT) of the RWST for adjustment of boron

concentration from 1 hour to 8 hours. Specifically, the minimum RWST

temperature requirement of TS 3.1.2.5b(3), TS 4.1.2.5b, and TS 4.5.5b

would be deleted. TS 3/4.1.2.6 would be revised as follows: (1) TS

3.1.2.6b, Action Statement b., and TS 4.1.2.6b, pertaining to the RWST,

would be deleted. (2) Editorial changes would be made to reflect the

deletion of the RWST requirements. TS 3/4.5.5 would be revised as

follows: the minimum RWST temperature requirement of TS 3.5.5c would be

deleted, and the action statement would be deleted and replaced with

two action statements. Action Statement a. would specify the

requirements when the RWST is inoperable due to boron concentration.

The action statement would also provide 8 hours to restore the boron

concentration to within the required limits. If boron concentration is

not restored within 8 hours, the action statement requires that the

unit be in hot standby within 6 hours and in cold shutdown within the

following 30 hours. Action Statement b. would specify the requirements

when the RWST is inoperable due to reasons other than boron

concentration. The associated Bases would also be appropriately

revised.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The increase in the RWST AOT does not alter the plant

configuration or operation. The potential for the RWST boron

concentration to be outside the TS limits is small because the RWST

and its contents are not involved with normal plant operation and

are not subject to process variations associated with plant

operation.

The potential causes of boron concentration deviation have been

evaluated with the conclusion that any deviation in RWST boron

concentration would not be expected to increase significantly during

the proposed 7 hour AOT increase.

The increase in the RWST AOT from 1 hour to 8 hours for reasons

directly related to boron concentration does not have a significant

effect on the accident analyses.

The removal of the redundant statement of RWST requirements from

TS 3.1.2.6 is an administrative change with no impact on plant

operation.

The removal of the minimum temperature limit for the RWST has no

effect on the plant configuration or operation. The removal of the

temperature limits does not affect any accident analyses since

evaluations have demonstrated that, due to the moderate climate at

DCPP, the RWST will not exceed the limits assumed in DCPP accident

analyses.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

Increasing the RWST AOT from 1 hour to 8 hours for reasons

directly related to boron concentration does not require physical

alteration to any plant system and does not change the method by

which any safety-related system performs its function.

The removal of the redundant statement of RWST requirements from

TS 3.1.2.6 is an administrative change that does not affect the

design and operation of the plant.

Deletion of the RWST temperature has no impact on any accident

analysis due to the moderate climate at DCPP. Additionally, the

deletion of the temperature does not require any physical alteration

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

performs its function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

of safety?

Increasing the RWST AOT for reasons directly related to boron

concentration does not affect any accident analysis assumptions,

initial conditions, or results. The margins of safety reflected in

the DCPP TS are not compromised by the 7 hour AOT increase.

Consequently, the proposed change does not have an effect on margin

of safety.

The removal of the redundant statement of RWST requirements from

TS 3.1.2.6 is an administrative change that does not affect the

requirements for the RWST nor alter its function.

The removal of the RWST temperature limits will not affect the

assumptions of any accident analysis because the moderate climate at

DCPP will prevent the temperature assumptions in the analyses from

being exceeded.

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 50.92(c) are

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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

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

NRC Project Director: Theodore R. Quay

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

323,Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County, California

Date of amendment request: August 17, 1994 (Reference LAR 94-07)

Description of amendment request: The proposed amendments would

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

Power Plant Unit Nos. 1 and 2 to relocate TS 3/4.4.2.1, ``Safety Valves

- Shutdown,'' 3/4.4.7, ``Chemistry,'' 3/4.4.9.2, ``Pressurizer

(Temperature Limits),'' 3/4.4.10, ``Structural Integrity,'' and 3/

4.4.11, ``Reactor Vessel Head Vents,'' in accordance with the

Commission's Final Policy Statement for relocation of current TS that

do not satisfy any of the screening criteria for retention. As part of

the relocation of TS 3/4.4.2.1, TS 3/4.4.2.2, ``Safety Valves -

Operating,'' would be revised to require that the pressurizer safety

valves be operable in Mode 4 with the reactor coolant system cold-leg

temperature greater than the low-temperature overpressure protection

system enable temperature, and TS 6.8, ``Procedures and Programs,''

would be revised to include the reactor coolant pump flywheel

inspection program. The specific TS changes proposed are as follows:

(1)

Technical Specifications (TS) 3/4.4.2.1 ``Safety Valves -

Shutdown,'' 3/4.4.7, ``Chemistry,'' 3/4.4.9.2, ``Pressurizer

(Temperature Limits),'' 3/4.4.10, ``Structural Integrity,'' 3/4.4.11,

``Reactor Vessel Head Vents,'' and TS 6.8, ``Procedures and Programs,''

would be revised in accordance with the Commission's Final Policy

Statement on TS Improvements for Nuclear Power Reactors.

(2)

TS 3/4.4.2.2, ``Safety Valves - Operating,'' would be revised to

require that the pressurizer safety valves be operable in Mode 4 with

the reactor coolant system cold-leg temperature greater than the low-

temperature overpressure protection system enable 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:

a. Do the changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes simplify the TS, meet regulatory

requirements for relocated TS, and implement the recommendations of

the Commission's Final Policy Statement on TS Improvements. Future

changes to these requirements will be controlled by 10 CFR 50.59.

The proposed changes are administrative in nature and do not involve

any modifications to any plant equipment or affect plant operation.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

b. Do the changes create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes are administrative in nature, do not

involve any physical alterations to any plant equipment, and cause

no change in the method by which any safety-related system performs

its function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

c. Do the changes involve a significant reduction in a margin of

safety?

The proposed changes do not alter the basic regulatory requirements

and do not affect any safety analyses. 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 50.92(c) are

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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

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

NRC Project Director: Theodore R. Quay

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

323,Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2,San Luis

Obispo County, California

Date of amendment request: August 17, 1994 (Reference LAR 94-09)

Description of amendment request: The proposed amendments would

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

Power Plant (DCPP) Unit Nos. 1 and 2 to change TS 3/4.4.9.1, ``Reactor

Coolant System - Pressure/Temperature Limits,'' Figures 3.4-2,

``Reactor Coolant System Heatup Limitations - Applicable Up to 8

EFPY,'' and 3.4-3, ``Reactor Coolant System Cooldown Limitations -

Applicable Up to 8 EFPY,'' to extend the applicability up to 12

effective full-power years (EFPYs). TS 3/4.4.9.3, ``Overpressure

Protection Systems,'' would be revised to specify a new low-temperature

overprotection (LTOP) system actuation pressure setpoint. The

associated Bases would also be appropriately revised. Additionally, TS

3/4.1.2.2, ``Flow Paths - Operating,'' TS 3/4.1.2.4, ``Charging Pumps -

Operating,'' TS 3/4.4.1.3, ``Hot Shutdown,'' TS 3/4.4.1.4.1, ``Cold

Shutdown - Loops Filled,'' TS 3/4.4.9.3, and TS 3/4.5.3, ``Tavg Less

than 350 Degrees F,'' would be revised to specify a new LTOP system

enable temperature.

(1) In TS 3/4.4.9.1, Figure 3.4-2, ``Reactor Coolant System Heatup

Limitations - Applicable Up to 8 EFPY,'' and Figure 3.4-3, ``Reactor

Coolant System Cooldown Limitations - Applicable Up to 8 EFPY,'' are

revised as follows:

(a) The ``Controlling Materials'' for the pressure/temperature

curves are revised to reflect the current reactor vessel beltline

region limiting weld and plate materials. (b)The title for the figures

is changed to reflect the applicability of the pressure/temperature

curves for up to 12 EFPYs of service life.

(2) The proposed changes to TS 3/4.4.9.3 are as follows:(a) The

LTOP enable temperature would be changed from 323 deg.F to 270 deg.F to

be consistent with Branch Technical Position (BTP) RSB 5-2, Revision 1,

Branch Position B.2.

(b) LTOP system actuation pressure setpoint would be revised from

less than or equal to 450 psig to less than or equal to 435 psig.

(3) TS 3/4.1.2.2, TS 3/4.1.2.4, TS 3/4.4.1.3, TS 3/4.4.1.4.1, TS 3/

4.4.9.3, and TS 3/4.5.3 would be revised to change the LTOP enable

temperature from 323 deg.F to 270 deg.F to be consistent with BTP RSB

5-2, Revision 1, Branch Position B.2. TS Bases 3/4.4.9.1 would be

revised to delete a reference to Table 4.4-5, ``Reactor Vessel Material

Surveillance Program - Withdrawal Schedule.'' The table was deleted

from the TS in Amendments 54 and 53 issued in July 1990. Reference to

the table in Bases 3/4.4.9 was inadvertently not deleted. The

information in this table is currently contained in the Final Safety

Analysis Report (FSAR) Update.

(4) TS Bases 3/4.4.9.3 would be revised to discuss limitations on

reactor coolant pump (RCP) and emergency core cooling system/chemical

and volume control system pump operation during low reactor coolant

system (RCS) temperature conditions.

(5) The other affected TS Bases would also be revised.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes to Figures 3.4-2 and 3.4-3 of TS 3.4.9.1

and the associated Bases will extend the applicability of the RCS

heatup and cooldown pressure/temperature limits from 8 to 12 EFPY.

Since the level of reactor vessel embrittlement projected for 12

EFPY is bounded by that previously projected for 8 EFPY, the

proposed changes will not impact the probability of brittle fracture

of the vessel, and consequently not impact the consequences of an

accident.

The present LTOP pressure setpoint was reviewed and found to be

acceptable and conservative for the extension of the pressure/

temperature curves to 12 EFPY. However, as a result of issues

unrelated to the change in the applicability of the pressure/

temperature curves, the LTOP actuation pressure setpoint is reduced.

The change accounts for pressure measurement error identified in NRC

IN [Information Notice] 93-58, a time delay in the LTOP system

actuation introduced as part of the installation of the Eagle 21

protection system, and additional conservatism incorporated into the

DCPP LTOP analysis. The changes to the pressure setpoint are

conservative and provide assurance that the maximum cold RCS

pressure will not be exceeded.

The proposed change to TS 3/4.1.2.2, 3/4.1.2.4, 3/4.4.1.3, 3/

4.4.9.3, and 3/4.5.3 will revise the LTOP enable temperature to be

consistent with the methodology and definition of ``low

temperature'' provided in BTP RSB 5-2 Revision 1. The proposed

changes do not involve physical alteration of the LTOP system or

change the method by which the LTOP system performs its function.

The proposed changes will benefit DCPP by expanding the RCS

pressure/temperature window, thereby increasing operator flexibility

during heatup and cooldown. This will decrease the probability of an

accident by decreasing the likelihood of an inadvertent PORV [power-

operated relief valve] actuation.

Deletion of reference to Table 4.4-5 from TS Bases 3/4.4.9 is

administrative in nature and does not affect plant operation.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes to TS 3.4.9.1 do not involve any physical

alteration to any plant system or change the method by which any

safety-related system performs its function. The probability of

catastrophic failure of the reactor vessel will not be changed as a

result of the extension of the curves to 12 EFPY.

The present LTOP pressure setpoint was reviewed and found to be

acceptable and conservative for the extension of the pressure/

temperature curves to 12 EFPY. However, as a result of issues

unrelated to the change in the applicability of the pressure/

temperature curves, the LTOP actuation pressure setpoint is reduced.

The change accounts for pressure measurement error identified in IN

93-58, a time delay in the LTOP system actuation introduced as part

of the installation of the Eagle 21 protection system, and

additional conservatism incorporated into the DCPP LTOP analysis.

The changes to the pressure setpoint are conservative and provide

assurance that the maximum cold RCS pressure will not be exceeded.

The proposed change to TS 3/4.1.2.2, 3/4.1.2.4, 3/4.4.1.3, 3/

4.4.9.3, and 3/4.5.3 will revise the LTOP enable temperature to be

consistent with the methodology and definitions provided in BTP RSB

5-2, Revision 1. Additionally, the proposed changes will not affect

the ability of the LTOP system to provide pressure relief at low

temperatures, thereby maintaining the LTOP design basis.

Deletion of reference to Table 4.4-5 from TS Bases 3/4.4.9 is

administrative in nature and does not result in physical alterations

or changes to the operation of the plant.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

of safety?

The proposed changes to TS 3.4.9.1 will extend the applicability

of the RCS heatup and cooldown pressure/temperature limits to 12

EFPY, but will not physically change these limits. The pressure/

temperature limits have been determined in accordance with 10 CFR

50, Appendix G, and include the safety margins with regard to

brittle fracture required by the ASME Code, Section III, Appendix G.

The RTndts determined for the reactor vessels at 12 EFPY are

lower than the values previously determined at 8 EFPY. Therefore,

there will be additional safety margin in the pressure/temperature

limits with respect to Appendix G requirements.

The change in the LTOP pressure setpoint is conservative and

provides assurance that the current margin of safety is maintained.

The proposed change to TS 3/4.1.2.2, 3/4.1.2.4, 3/4.4.1.3, 3/

4.4.9.3, and 3/4.5.3, will revise the LTOP enable temperature to be

consistent with the methodology and definitions provided in BTP RSB

5-2, Revision 1, which provides the requirements for reactor vessel

overpressurization protection at low temperatures.

Deletion of reference to Table 4.4-5 from TS Bases 3/4.4.9 is an

administrative change and does not involve any physical alteration

to the plant.

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 50.92(c) are

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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

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

NRC Project Director: Theodore R. Quay

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,

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

Pennsylvania

Date of amendment request: August 12, 1994

Description of amendment request: The amendment would revise the

Limiting Condition for Operation for the Emergency Core Cooling System

specified in Technical Specifications Section 3.5.1 and associated

Bases Section 3.4.5.1 to include a new ACTION statement in the event

that the High Pressure Coolant Injection system and one Core Spray

subsystem, and/or one Low Pressure Coolant Injection subsystem, are

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:

1. The proposed Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS change does not involve any physical changes to

plant systems or components, nor does it affect the ability of the

Low pressure Coolant Injection (LPCI) Core Spray (CS), and High

Pressure Coolant Injection (HPCI) systems to respond to an accident.

These systems are not accident initiators, since their design

function is accident mitigation.

This proposed TS change, which only addresses equipment status,

will not significantly increase the probability of occurrence of an

accident previously evaluated. The addition of the proposed ACTION

statement enables the plant not to implement TS Section 3.0.3, which

requires a plant shutdown, when the HPCI system is inoperable in

conjunction with one (1) CS subsystem, and/or one (1) LPCI

subsystem. The proposed TS change does not impact the operation of

any equipment important to safety. This proposed TS change does not

make physical modifications to the plant or to equipment, nor does

it impact any design requirements of the HPCI, CS, and LPCI systems.

The proposed TS change does not introduce any failure mechanisms of

a different type than those previously evaluated, since no physical

changes are being made to the facility. This proposed change will

not create any new failure modes which would cause plant equipment

to malfunction more frequently than previously evaluated.

The basis for TS Sections 3.8.2.1 and 3.8.3.1, which specify

that four (4) independent divisions of Safeguard dc electrical power

shall be operable, or shall be restored to operability with 8 hours,

is to ensure that sufficient power is available to supply safety-

related equipment required to safely shut down the plant, and to

provide for mitigation and control of accident conditions at the

plant. As discussed in Section 6.3.2 of the NRC Safety Evaluation

Report (SER), i.e., NUREG-0991, ``Safety Evaluation Report Related

to the Operation of Limerick Generating Station, Units 1 and 2,''

dated August 1983, the most limiting single failure for the

Emergency Core Cooling System (ECCS), which includes all break

sizes, is the failure of the dc power system common to the HPCI

system, one (1) CS subsystem, and one (1) LPCI subsystem. Only one

(1) single failure is assumed to occur in the event of a Design

Basis Accident (DBA). Therefore, three (3) LPCI pumps, one (1) CS

subsystem, and the Automatic Depressurization (ADS) system would be

operable and available, for use in the event of a DBA, to provide

sufficient core cooling to safely shut down the plant. Although the

loss of Division 2 dc power specifically impacts the ``B'' LPCI and

``B'' CS, the analysis performed in the NRC SER evaluates the number

of ECCS available for use in a DBA. Since the amount of available

core cooling is independent of which loop of LPCI or CS is assumed

to fail, this analysis is applicable to the loss of any division/

loop of LPCI or CS. Therefore, the loss of the HPCI system, one (1)

CS subsystem, and/or one (1) LPCI subsystem is bounded by the

existing analysis. Since the loss of HPCI, one (1) CS subsystem,

and/or one (1) LPCI subsystem is an analyzed condition, and actions

associated with TS Section 3.0.3 are related to unanalyzed

conditions, the requirements of TS Section 3.0.3 are not applicable

to this scenario. Adding an ACTION statement, as proposed, identical

to the ACTION statement which currently applies to the loss of

Division 2 of Safeguard dc electrical power causes no change in the

consequences of any accidents previously evaluated. This proposed TS

change does not impact systems, structures, and components designed

to mitigate the consequences of an accident. In the event of an

accident, the plant configuration following the event will be within

the bounds of the existing analysis, and there will be no change in

the radiological consequences due to an accident.

Therefore, the proposed TS change does not involve an increase

in the probability or consequences of an accident [previously]

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed TS change does not require any physical changes to

plant systems or equipment, nor will it affect the ability of the

HPCI, CS, and LPCI systems from performing their design functions,

which is to mitigate the consequences of an accident. These systems

do not contribute to the initiation of an accident, since their

function is accident mitigation. This proposed TS change will not

introduce new equipment malfunction or failure modes. The proposed

TS change will not introduce any failure mechanisms of a different

type than those previously evaluated. The existing design basis for

the plant, as described in Section 6.3.2.5 of the LGS Updated Safety

Analysis Report (UFSAR) and Section 6.3.2 of the NRC SER, bounds the

condition proposed by this TS Change Request. Section 6.3.2 of the

NRC SER indicates that the most limiting single failure for the ECCS

is the loss of the dc system powering the HPCI, CS, and LPCI

systems. Assuming this failure, three (3) LPCI pumps, one (1) CS

subsystem, and the ADS would still be operable and available, for

use in the event of a DBA, to ensure adequate core cooling to safely

shut down the plant. Although the loss of Division 2 dc power

specifically affects ``B'' LPCI and ``B'' CS, the analysis performed

in the NRC SER evaluates the number of ECCS available for use in a

DBA. Since the amount of available core cooling is independent of

which loop of LPCI or CS is assumed to fail, this analysis is

applicable to the loss of any division/loop of LPCI or CS. Since the

loss of HPCI, one (1) CS subsystem, and/or one (1) LPCI subsystem,

is an analyzed condition, and the actions associated with TS Section

3.0.3 pertain to unanalyzed conditions, the requirements of TS

Section 3.0.3 do not apply to the condition proposed by this TS

Change Request.

Therefore, the proposed TS change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

The proposed TS change [TS] does not involve any physical

changes to the design or functional requirements of the LPCI, CS, or

HPCI systems. These systems will continue to function as designed to

mitigate the consequences of an accident.

This proposed TS change involves adding an additional ACTION

statement, and revising the associated supporting Bases section, to

specifically address the inoperability of the HPCI system in

conjunction with the inoperability of one (1) CS subsystem, and/or

one (1) LPCI subsystem. These systems would be inoperable in the

event of the loss of Division 2 of the Safeguard dc electrical power

supply. The Bases associated with Safeguard electrical power

systems, which provide power to equipment required to safely

shutdown the plant and to mitigate consequences of an accident, are

unchanged. The proposed TS change involves adding an ACTION

statement which is identical to the ACTION statement which addresses

the inoperability of Division 2 of Safeguard dc power, which is a

condition analyzed in the LGS UFSAR and NRC SER. Therefore, the

proposed TS change to include an additional ACTION statement does

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, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Mohan C. Thadani, Acting

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,

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

Pennsylvania

Date of amendment request: August 23, 1994

Description of amendment request: This amendment would remove the

125/250 Vdc Class 1E Battery Load Cycle Table from Technical

Specifications, which is consistent with NUREG-1433, ``Standard

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. The proposed TS change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

This proposed change removes the repetitious 125/250 Vdc Class

1E Battery Load Cycle Table which is also found in the LGS Updated

Final Safety Analysis Report (UFSAR). The proposed change to TS does

not affect the requirement to perform surveillance testing and the

manner of performing surveillance testing is adequately described in

plant procedures. The UFSAR containing the Battery Load Cycle Table

and station procedures are maintained using the provisions of 10 CFR

50.59 and are subject to the change control process in the

Administrative Controls Section of the LGS TS Section 6.0. Since any

future changes to these controlled documents will be evaluated per

10 CFR 50.59, no [changes] (significant or insignificant) in the

probability or consequences of an accident previously evaluated will

be allowed. Therefore, this change will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

This proposed change removes the repetitious 125/250 Vdc Class

1E Battery Load Cycle Table which is also found in the LGS Updated

Final Safety Analysis Report (UFSAR). This change will not alter the

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

installed) or make changes to methods governing normal plant

operations. This change will not impose different requirements and

adequate control of information will be maintained. The manner of

performing surveillance testing can be adequately described in plant

procedures. The proposed change will remove the table, and will not

alter assumptions made in the safety analysis and licensing basis.

Therefore, this change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

This proposed change removes the repetitious 125/250 Vdc Class

1E Battery Load Cycle Table which is also found in the LGS Updated

Final Safety Analysis Report (UFSAR). The change will not reduce the

margin of safety since the location of the Battery Table has no

impact on any safety analysis assumptions. Since all Battery Load

Table changes (i.e., UFSAR Changes) and procedure changes are

evaluated per the requirements of 10 CFR 50.59, no reduction

(significant or insignificant) in the margin of safety will be

allowed. Therefore, this change will not involve a significant

reduction in a margin of safety.

The existing requirements for NRC review and approval of

revisions, in accordance with 10 CFR 50.90, to those details and

requirements proposed for deletion, do not have a specific margin of

safety upon which to evaluate. However, since the proposed change is

consistent with the BWR Standard Technical Specifications (NUREG-

1433), revising the TS to reflect the approved level of detail and

requirements ensures no reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Mohan C. Thadani, Acting

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: August 5, 1994

Description of amendment request: The proposed amendment

incorporates line item Technical Specification improvements listed in

Generic Letter 93-05 relevant to Emergency Diesel Generator (EDG)

surveillance requirements. The proposed amendment eliminates the

requirements to start EDGs with an inoperable offsite circuit(s) of AC

electrical power and adds a provision that eliminates required testing

of the remaining EDGs when one EDG is inoperable due to an inoperable

support system or an independently testable component with no potential

for common mode failure for the remaining EDGs. In addition, if testing

of the EDGs is required, then the surveillances will be performed

within 16 hours instead of 24 hours as currently specified.

The proposed amendment also deletes the requirement to perform a

loss of offsite power (LOP) test following the 24-hour EDG endurance

run test. In its place, a hot restart test (no LOP load sequencing)

will be established.

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.

LCR 94-10

The proposed changes in this License Change Request (LCR) have

been extensively reviewed by the NRC during the preparation of

NUREG-1366 and Generic Letter 93-05, and by [Public Service Electric

and Gas Company] PSE&G during the development and approval of this

LCR. The LCR revises the current ACTION statement of Technical

Specification 3.8.1.1 to eliminate testing of the unaffected

Emergency Diesel Generators (EDGs) upon loss of an offsite power

circuit(s) and/or an EDG. The basis for this testing was originally

to verify the reliability of the EDGs, however, as stated in NUREG-

1366, industry experience has shown that excessive testing of the

EDGs has in fact reduced reliability.

The EDG design and function remain as previously analyzed and

the EDG response during accident conditions is not affected. This

change will improve EDG performance by reducing the number of

unnecessary starts and by requiring more appropriate testing (within

16 hours instead of 24 hours) when there is a potential common mode

failure.

These changes will not result in a significant increase in the

probability or consequences of a previously evaluated accident, nor

will it result in a significant reduction in a margin of safety.

LCR 94-13

The proposed changes in this License Change Request (LCR) have

been extensively reviewed by the NRC during the preparation of

NUREG-1366 and Generic Letter 93-05, and by PSE&G during the

development and approval of this LCR. Regulatory Guide 1.108, Rev.

1, states that the performance of a loss of Off-site Power (LOP)

test (Surveillance Requirement 4.8.1.1.2.h.4.b) immediately

following the 24 hour endurance run demonstrates that the Emergency

Diesel Generator (EDG) can start in the prescribed time when the EDG

is at its normal operating temperature. The purpose of performing

the LOP test immediately following the 24 hour endurance run is to

demonstrate the hot restart capability of the EDG at full load

conditions. However, demonstrating diesel generator hot restart

capability without loading the engine does not invalidate or reduce

the effectiveness of the hot restart test. Performance of this test

can be conducted in any plant condition since its performance at

power will have no adverse effect on plant operations.

The LOP test will continue to be performed at standby conditions

to provide assurance that the EDG is capable of responding to a LOP

as assumed in the accident analyses.

EDG design and function remain as previously analyzed. Their

response during accident conditions [is] not affected by these

changes. Therefore, no significant increase in the probability of an

accident previously evaluated results from these changes.

2. Will not create the possibility of a new or different kind of

accident from any accident previously evaluated.

LCR 94-10

The elimination of the unnecessary EDG starts will not result in

any change in plant configuration or operation. Therefore, the

proposed changes will not create the possibility of a new or

different kind of accident from any accident previously evaluated or

analyzed.

LCR 94-13

The proposed revisions to the Technical Specifications do not

involve a physical change in any system configuration and do not

introduce new operating configurations. These changes will not

result in any net reduction in testing and will not affect EDG

reliability. This test may be performed in any plant condition since

its performance at power will have no adverse effect on plant

operations. Therefore, these changes do not create the possibility

of a new or different kind of accident from any previously

evaluated.

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

safety.

LCR 94-10

The changes proposed in this LCR do not reduce the ability of

any system or component to perform its safety related function. The

basis of NUREG-1366, Generic Letter 93-05 and the analysis performed

in support of this LCR is that the reduction in unnecessary EDG

starts can improve safety by diminishing challenges to plant systems

and reducing equipment wear or degradation. These proposed changes

involve only surveillance frequencies and do not change the method

of performing any surveillance. The operation of systems and

equipment remains unchanged. Therefore, eliminating unnecessary EDG

starts does not involve a reduction in the margin of safety.

LCR 94-13

Surveillance testing per the proposed Technical Specifications

would continue to demonstrate the ability of the EDGs to perform

their intended function of providing electrical power to the

emergency safety systems needed to mitigate design basis transients

consistent with the plant safety analyses. The margin of safety

demonstrated by the plant safety analyses is therefore not affected

by the proposed changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and

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

NRC Project Director: Mohan C. Thadani, Acting

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

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: August 19, 1994

Description of amendment request: The proposed changes add a new

statement (b) to Limiting Condition for Operation (LCO) 3.1.3.2.1, Rod

Position Indication Systems, and reletters the existing action

statement (b) to (c). The new action (b) will read:

With two or more analog rod position indicators per bank

inoperable, within one hour restore the inoperable rod position

indicator(s) to OPERABLE status or be in Hot Standby within the next

6 hours. A maximum of one rod position indicator per bank may remain

inoperable following the one hour, with Action (a) above being

applicable from the original entry time into the LCO.

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.

The request (both proposed changes) does not change any

assumption or parameter assumed to function in any of the design/

licensing basis analysis, and therefore the probability or

consequences of an accident previously evaluated are not increased.

The change, as described in section IB, [the addition of the new

action statement] incorporates into the applicable LCO the action

statement which is already taken under technical specification

3.0.3, and does not alter the operator response or response time.

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

accident from any accident previously evaluated.

The proposed change does 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 change

does not change any assumption or parameter assumed to function in

any of the design/licensing basis analysis. No changes to the

operator response or operator response time is proposed, only that

the response is now taken under the confines of the LCO.

Therefore, there is no reduction in any margin of safety from

the proposed changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: Mohan C. Thadani, Acting

South Carolina Electric & Gas Company, South Carolina Public

ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear

Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: July 20, 1994, as supplemented September

20, 1994

Description of amendment request: The proposed change would modify

the Virgil C. Summer Nuclear Station, Unit 1, (VCSNS) Technical

Specifications (TS) to allow alternative, equivalent testing of diesel

fuel used in the emergency diesel generators (EDG). These alternative

methods are necessary due to recent changes in Environmental Protection

Agency (EPA) Regulations that are designed to limit the use of high

sulfur fuels. The licensee also proposes to modify the VCSNS TS by

changing the revision level of WCAP-10216-P-A, ``Relaxation of Constant

Axial Offset Control - FQ Surveillance Technical Specification,''

referenced in TS 6.9.1.11. This pertains to the FQ(z) TS (TS 3.2.1 and

3.2.2) and is necessary since Westinghouse revised their methodology in

determining FQ(z).

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 probability or consequences of an accident previously

evaluated is not significantly increased.

The change in testing methods for the EDG fuel oil has no impact

on the probability or consequences of any design basis accident.

These tests have been determined to be equivalent to the previously

approved testing methods and are needed due to changes in the EPA's

regulations regarding sulfur in motor vehicle fuels. The dye used to

identify high sulfur fuels will have no adverse affect on the

performance of the EDG's. The proposed testing assures a continued

high level of quality of the diesel fuel received and stored on

site.

The change in revision level of a reference in TS section

6.9.1.11 has no impact on the probability of occurrence or

consequences of any design basis accident. All design and

performance criteria will continue to be met and no new single

failure mechanisms will be created. The change in revision level for

WCAP-10216-P-A does not involve any alterations to plant equipment

or procedures which could affect any operational modes or accident

precursors. This change only incorporates by reference, the

methodology for determining the penalty to be used in calculating

Core Operating Limits. This methodology allows the penalty to be

cycle specific and is primarily affected by the core configuration.

This penalty is used for normal operation and provides more

conservatism to the core operation for the cycle.

2. [The proposed license amendment does not] create the

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

previously evaluated.

The change in testing methods for the EDG fuel oil will not

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

any accident previously evaluated. These tests have been determined

by the EPA and other organizations to be equivalent to the

previously approved testing methods. The effect of the blue dye,

used to identify high sulfur fuels, on the performance of the EDGs

has been evaluated and determined to be insignificant. The testing

proposed assures a continued high level of quality for the diesel

fuel received and stored on site.

The change of revision level of a reference in TS section

6.9.1.11 has no impact on the probability of occurrence or

consequences of any design basis accident. All design and

performance criteria will continue to be met and no new single

failure mechanisms will be created. The change in revision level for

WCAP-10216-P-A does not involve any alterations to plant equipment

or procedures which could affect any operational modes or accident

precursors. This change only incorporates, by reference, the

methodology for determining the penalty to be used in calculating

Core Operating Limits. This methodology allows the penalty to be

cycle specific and is primarily affected by the core configuration.

This penalty is used for normal operation and provides more

conservatism to the core operation for the cycle.

3. [The proposed license amendment does not] involve a

significant reduction in a margin of safety.

The change in testing methods for the EDG fuel oil will not

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

testing methods have been determined to be equivalent to the

previously approved testing methods. The test for sulfur assures

that the sulfur content is within the allowable range for weight-

percent. The test for color and clarity assures that the fuel is

relatively free of water and particulate contaminants. The proposed

tests provide at least an equivalent level of quality and

repeatability for the fuel oil analysis, thus assuring that the

margin of safety is not reduced.

The change in revision level of a reference in TS section

6.9.1.11 does not change the proposed reload design or safety

analysis limits for each cycle reload core. The associated change to

WCAP-10216-P-A due to the revision will be specifically evaluated

using approved reload design methods. The larger penalty actually

provides for an increase in margin during certain burnup ranges.

Since the safety analysis limits are unaffected, and the cycle

specific analysis will show that the analysis limits are met, the

change proposed will have no adverse impact on 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: Fairfield County Library,

Garden and Washington Streets, Winnsboro, South Carolina 29180

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218

NRC Project Director: David B. Matthews

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

50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,

San Diego County, California

Date of amendment requests: July 28, 1994

Description of amendment requests: The licensee proposes revisions

to Technical Specification (TS) 3.9.8.1, ``Shutdown Cooling and Coolant

Circulation -- High Water Level,'' TS 3.9.8.2, ``Shutdown Cooling and

Coolant Circulation -- Low Water Level,'' and their Bases to facilitate

testing of low-pressure safety injection system components and permit

additional flexibility in scheduling maintenance on the shutdown

cooling system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No

Limiting Conditions for Operation (LCO) in Technical

Specifications (TSs) 3.9.8.1 and 3.9.8.2 define the operability

requirements for the Shutdown Cooling (SDC) system during refueling

operations (Mode 6) while the water level above the top of the

reactor vessel flange is at least 23 feet and less than 23 feet,

respectively. The objective of these TSs is to ensure that (1)

sufficient cooling is available to remove decay heat, (2) the water

in the reactor vessel is maintained below 140 degrees Fahrenheit,

and (3) sufficient coolant circulation is maintained in the reactor

core to minimize boron stratification leading to a boron dilution

incident.

The proposed TS changes affect the current limits imposed while

ensuring adherence to the basis of the TS. No plant modifications

are being made. The reactor cavity water level limitations and SDC

system required operating times are being changed based on plant

specific calculations and the objectives of the TSs are being

maintained.

(1) reduce the water level where two trains of SDC are required

from 23 feet to 20 feet above the reactor pressure vessel flange,

In the Bases Section 3/4.9.8, it is stated that ``With the

reactor vessel head removed and 23 feet of water above the reactor

pressure vessel flange, a large heat sink is available for core

cooling, thus in the event of a failure of the operating shutdown

cooling train, adequate time is provided to initiate emergency

procedures to cool the core.''

In the Bases for the New Standard Technical Specifications,

``NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4

it is stated that; ``The 23 ft level was selected because it

corresponds to the 23 ft requirement established for fuel movement

in LCO 3.9.6, ``Refueling Water Level.''

Southern California Edison (Edison) calculations show that there

is a minimal difference in the time to boil due to the 3-foot change

in required water level. Therefore, adequate water is still

available to mitigate the consequences of losing SDC.

(2) increase the time a required train of the SDC system may be

removed from service from up to 1 hour per 8-hour period to up to 2

hours per 8-hour period,

(3) allow the SDC system to be removed from service to allow

testing of Low Pressure Safety Injection system components,

The proposed TS changes the time the SDC train may be removed

from operation from up to 1 hour per 8-hour period to up to 2 hours

per 8-hour period, and allows removal of the SDC train from

operation for testing of the Low Pressure Safety Injection (LPSI)

system components as well as for core alterations in the vicinity of

the hot legs. The proposed TS change also imposes certain

restrictions to ensure operating the SDC system in accordance with

this proposed TS change is of no safety significance. These

restrictions are discussed separately below.

When securing the only operating train of the SDC system, the

maximum Reactor Coolant System (RCS) temperature is maintained less

than or equal to 140 degrees Fahrenheit. The initial conditions and

heatup rate are selected such that the RCS temperature remains less

than or equal to 140 degrees Fahrenheit during the test. Therefore,

there is ample margin to boiling. Typical initial temperatures are

less than 100 degrees Fahrenheit.

The water being injected by the LPSI system test is cool water

from the Refueling Water Storage Tank (RWST) and will increase the

available inventory providing the heat sink by several inches. The

two hours is sufficient time to align the system to test, perform

the test, and restore the train of SDC to operation prior to

exceeding 140 degrees Fahrenheit.

No operations are permitted that would cause a reduction of the

RCS boron concentration. This minimizes the probability of an

inadvertent boron dilution event. The use of adequately borated

water for injection into the RCS during the test provides assurance

that the test itself cannot lead to a boron dilution event. When the

SDC system is operating, the minimum SDC flow rate of 2200 gpm

imposed by TS 4.9.8.1 and TS 4.9.8.2 is sufficient to ensure

complete mixing of the boron within the RCS.

The LPSI component testing is only allowed when the reactor

cavity water level is maintained greater than or equal to 20 feet

above the reactor pressure vessel flange. This level ensures an

adequate heat sink to perform the LPSI pump suction header check

valve test.

(4) allow for running 1 train of shutdown cooling with

additional requirements when the water level is less than 20 feet

but greater than 12 feet above the reactor pressure vessel flange,

(5) add an action to be taken when operating 1 train of SDC with

less than 20 feet above the reactor pressure vessel flange when the

specified requirements are not met,

In the event of a loss of SDC, the time to boil is reduced from

approximately 3.7 hours when the water level is 23 feet above the

reactor vessel flange to approximately 2.3 hours at 12 feet,

assuming the reactor has only been shutdown for 6 days. However,

this is ample time to close containment (less than 1 hour) and to

restore SDC or initiate alternative cooling (e.g., add water to the

cavity (approximately 1 hour)). Twelve feet of water above the

reactor vessel flange corresponds to 24 feet 8-7/8 inches above the

active fuel.

Requiring the reactor to be shutdown for at least 6 days to have

only one train of SDC operable when the reactor cavity level is

between 20 feet and 12 feet above the reactor pressure vessel flange

ensures that the time to boil is greater than twice the time it

would take us to establish containment closure and to commence

reactor cavity fill with the required standby equipment.

One train of SDC operating with a containment spray pump allows

for the high capacity LPSI pump to be the main standby pump capable

of filling the reactor cavity to at least 20 feet above the reactor

pressure vessel flange upon loss of SDC. The high pressure safety

injection pump will also be maintained ready to increase the water

level if needed. In support of this contingency the RWST will be

required to contain the volume of water required to raise the level

to 20 feet above the reactor vessel flange.... The reactor cavity

can be filled at a rate of approximately 4.0 inches per minute with

the LPSI pump.

If operating one train of the SDC system with less than 20 feet

of water above the reactor pressure vessel flange and any of the

required conditions are not met, requiring immediate action to

establish greater than or equal to 20 feet of water above the

reactor pressure vessel flange ensures no time is wasted trying to

restore conditions that should be used to increase the volume of

water of the heat sink. By taking action to restore the level to 20

feet above the reactor pressure vessel flange the plant will be

placed in TS 3.9.8.1, which only requires one train of SDC to be

operable. Additionally, the core will not heat up while the water

level in the reactor cavity is being raised with cool water from the

RWST. This will provide additional time to either restore the one

train of SDC or take other actions to provide core cooling.

A Probabilistic Risk Assessment (PRA), with (a) one train of the

SDC system operable with the reactor cavity water level greater than

or equal to 12 feet above the reactor pressure vessel flange, and

(b) one train of the SDC system operable with the reactor cavity

water level greater than or equal to 20 feet above the reactor

pressure vessel flange, showed that the operations in accordance

with the proposed TS would not significantly increase the

probabilities of inventory boiling and core damage.

(6) delete the obsolete reference to the implementation of DCP

2-6863 and MMP 3-6863,

This is an editorial change.

(7) delete an obsolete footnote allowing removal of both trains

of SDC with the water less than 23 feet above the reactor vessel

flange from the Unit 3 TSs.

This is an editorial change.

Therefore, proposed changes 1 through 7 do not involve a

significant increase in the probability or consequences of an

accident.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No

(1) reduce the water level where two trains of SDC are required

from 23 feet to 20 feet above the reactor pressure vessel flange,

(2) increase the time a required train of the SDC system may be

removed from service from up to 1 hour per 8-hour period to up to 2

hours per 8-hour period,

(3) allow the SDC system to be removed from service to allow

testing of Low Pressure Safety Injection system components,

(4) allow for running 1 train of shutdown cooling with

additional requirements when the water level is less than 20 feet

but greater than 12 feet above the reactor pressure vessel flange,

(5) add an action to be taken when operating 1 train of SDC with

less than 20 feet above the reactor pressure vessel flange when the

specified requirements are not met,

The Limiting Conditions for Operation (LCO) in Technical

Specifications (TSs) 3.9.8.1 and 3.9.8.2 define the operability

requirements for the SDC system during refueling operations (Mode 6)

while the water level above the top of the reactor vessel flange is

at least 23 feet and less than 23 feet, respectively. The objective

of the proposed TS changes is to ensure that the intent of the Bases

is maintained. [i.e., (1) sufficient cooling is available to remove

decay heat, (2) water in the reactor vessel is maintained below 140

degrees Fahrenheit, and (3) sufficient coolant circulation is

maintained in the reactor core to minimize boron stratification

leading to a boron dilution incident.]

The proposed TS changes affect the current limits imposed while

ensuring adherence to the basis of the TS. No plant modifications

are being made. The reactor cavity water level limitations and SDC

system required operating times are being changed based on plant

specific calculations and the objective of the TSs are being

maintained. The added requirements and action statement facilitate

safe operation.

(6) delete the obsolete reference to the implementation of DCP

2-6863 and MMP 3-6863, and

This is an editorial change.

(7) delete an obsolete footnote allowing removal of both trains

of SDC with the water less than 23 feet above the reactor vessel

flange from the Unit 3 TSs.

This is an editorial change.

Therefore, the operation of the facility in accordance with

proposed changes 1 through 7 does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No

Limiting Conditions for Operation (LCO) in TSs 3.9.8.1 and

3.9.8.2 define the operability requirements for the SDC system

during refueling operations (Mode 6) while the water level above the

top of the reactor vessel flange is at least 23 feet and less than

23 feet, respectively. The objective of these TSs is to ensure that

(1) sufficient cooling is available to remove decay heat, (2) the

water in the reactor vessel is maintained below 140 degrees

Fahrenheit, and (3) sufficient coolant circulation is maintained in

the reactor core to minimize boron stratification leading to a boron

dilution incident.

(1) reduce the water level where two trains of SDC are required

from 23 feet to 20 feet above the reactor pressure vessel flange,

In the Bases Section 3/4.9.8, it is stated that ``With the

reactor vessel head removed and 23 feet of water above the reactor

pressure vessel flange, a large heat sink is available for core

cooling, thus in the event of a failure of the operating shutdown

cooling train, adequate time is provided to initiate emergency

procedures to cool the core.''

In the Bases for the New Standard Technical Specifications,

``NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4

it is stated that: ``The 23 ft level was selected because it

corresponds to the 23 ft requirement established for fuel movement

in LCO 3.9.6, ``Refueling Water Level.''

Edison calculations show that there is a minimal difference in

the time to boil due to the 3-foot change in required water level.

Therefore, the margin of safety has not been significantly reduced.

(2) increase the time a required train of the SDC system may be

removed from service from up to 1 hour per 8-hour period to up to 2

hours per 8-hour period,

(3) allow the SDC system to be removed from service to allow

testing of Low Pressure Safety Injection system components,

The proposed TS changes the time the SDC train may be removed

from operation from up to 1 hour per 8-hour period to up to 2 hours

per 8-hour period, and allows removal of the SDC train from

operation for testing of the LPSI system components as well as for

core alterations in the vicinity of the hot legs. The proposed TS

change also imposes certain restrictions to ensure operating the SDC

system in accordance with this proposed TS change is of no safety

significance. These restrictions are discussed separately below.

When securing the only operating train of the SDC system, the

maximum RCS temperature is maintained less than or equal to 140

degrees Fahrenheit. The initial conditions and heatup rate are

selected such that RCS temperature remains less than or equal to 140

degrees Fahrenheit during the test. Therefore, there is ample margin

to boiling. Typical initial temperatures are less than 100 degrees

Fahrenheit.

The water being injected by the LPSI system test is cool water

from the RWST and will increase the available inventory providing

the heat sink by several inches. The two hours is sufficient time to

align the system to test, perform the test, and restore the train of

SDC to operation prior to exceeding 140 degrees Fahrenheit.

No operations are permitted that would cause a reduction of the

RCS boron concentration. This minimizes the probability of an

inadvertent boron dilution event. The use of adequately borated

water for injection into the RCS during the test provides assurance

that the test itself cannot lead to a boron dilution event. When the

SDC system is operating, the minimum SDC flow rate of 2200 gpm is

sufficient to ensure complete mixing of the boron within the RCS.

The LPSI component testing is only allowed when the reactor

cavity water level is maintained greater than or equal to 20 feet

above the reactor pressure vessel flange. This level ensures an

adequate heat sink to perform the LPSI pump suction header check

valve test.

The added requirements and the nature of the test provide

assurances that the water temperature will be maintained less than

140 degrees Fahrenheit and that boron stratification is prevented.

(4) allow for running 1 train of shutdown cooling with

additional requirements when the water level is less than 20 feet

but greater than 12 feet above the reactor pressure vessel flange,

(5) add an action to be taken when operating 1 train of SDC with

less than 20 feet above the reactor pressure vessel flange when the

specified requirements are not met,

In the event of a loss of SDC, the time to boil is reduced from

approximately 3.7 hours at 23 feet to approximately 2.3 hours at 12

feet, when the reactor has only been shutdown for 6 days. However,

this is ample time to close containment (less than 1 hour), and to

restore SDC or initiate alternative cooling (e.g., add water to the

cavity (approximately 1 hour)).

Requiring the reactor to be shutdown for at least 6 days to have

only one train of SDC operable when the reactor cavity level is

between 20 feet and 12 feet above the reactor pressure vessel flange

ensures that the time to boil is greater than twice the time it

would take us to establish containment closure and to commence

reactor cavity fill with the required standby equipment.

One train of SDC operating with a containment spray pump allows

for the high capacity LPSI pump to be the main standby pump capable

of filling the reactor cavity to at least 20 feet above the reactor

pressure vessel flange upon loss of SDC. The high pressure safety

injection pump will also be maintained ready to increase the water

level if needed. In support of this contingency the RWST will be

required to contain the volume of water required to raise the level

to 20 feet above the reactor vessel flange. The reactor cavity can

be filled at a rate of approximately 4.0 inches per minute with the

LPSI pump.

If operating one train of the SDC system with less than 20 feet

of water above the reactor pressure vessel flange and any of the

required conditions are not met, requiring immediate action to

establish greater than or equal to 20 feet of water above the

reactor pressure vessel flange ensures no time is wasted trying to

restore conditions that should be used to increase the volume of

water of the heat sink. By taking action to restore the level to 20

feet above the reactor pressure vessel flange the plant will be

placed in TS 3.9.8.1, which only requires one train of SDC to be

operable. Additionally, the core will not heat up while the reactor

cavity water level is being raised with cool water from the RWST.

This will provide additional time to either restore the one train of

SDC or take other actions to provide core cooling.

A PRA showed that the operations in accordance with the proposed

TS did not significantly increase the probabilities of inventory

boiling and core damage.

(6) delete the obsolete reference to the implementation of DCP

2-6863 and MMP 3-6863,

This is an editorial change.

(7) delete an obsolete footnote allowing removal of both trains

of SDC with the water less than 23 feet above the reactor vessel

flange from the Unit 3 TSs.

This is an editorial change.

Therefore, operation of the facility in accordance with proposed

changes 1 through 7 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 50.92(c) are

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

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

Attorney for licensee: James A. Beoletto, Esquire, Southern

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

NRC Project Director: Theodore R. Quay

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

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

Date of amendment request: September 9, 1994 (TS 94-04)

Description of amendment request: The proposed change would revise

specifications associated with the cold leg accumulators (CLAs).

Specifically, the proposed amendment would: (1) remove a footnote from

Specification 3.5.1.1.c that applied to Unit 2 Cycle 6 operation only;

(2) add a requirement to Specification 3.5.1.1 that power be removed

from the CLA isolation valve when the reactor coolant system pressure

is above 2000 psig; (3) modify Specification 3.5.1.1 Action Statement

a. to indicate that with a CLA inoperable for reasons other than the

boron concentration not being within limits, the CLA must be returned

to operable status within 1 hour or the plant placed in the hot standby

condition, and the pressurizer pressure reduced to 1000 psig or less

within the next 6 hours; (4) modify Specification 3.5.1.1 Action

Statement b. to indicate that with a CLA inoperable because the boron

concentration is not within limits, the boron concentration must be

restored to within limits within 72 hours or the plant placed in the

hot standby condition within the next 6 hours and the pressurizer

pressure reduced to 1000 psig or less within the next 6 hours; (5)

remove the wording from Specification 4.5.1.1.1.a.1 for using the

absence of alarms or level measurement as the technique used to verify

CLA volume and pressure; (6) add the requirement to Specification

4.5.1.1.1.a.2 to verify that the CLA isolation valve is ``fully open''

rather than ``open;'' (7) modify Specification 4.5.1.1.1.b to show that

verification of boron concentration is not required for additions from

the refueling water storage tank, and add a footnote to indicate that

the verification is required only if the affected accumulator

experienced a volume increase; (8) modify Specification 4.5.1.1.1.c to

show that the test is satisfied by verifying that power is removed from

the isolation valve, not that the valve operator is disconnected by

removal of the breaker from the circuit; (9) delete Specification

4.5.1.1.1.d to verify that each CLA isolation valve opens automatically

when reactor coolant pressure exceeds the P-11 setpoint, and upon

receipt of a safety injection signal; (10) delete Specification

4.5.1.1.2 to verify the accumulator water level and pressure channels

operable by performing Channel Functional and Calibration tests, and

delete the related footnote; (11) change ``tanks'' to ``each cold leg

injection accumulator;'' and (12) revise the associated Bases where

necessary to reflect these 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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to TS 3.5.1.1 implement revised action

times for cold leg injection accumulator (CLA) inoperability.

Several other clarifications and enhancements have been incorporated

to provide consistency with the latest version of standard TSs

(NUREG-1431). The new action times provide a prompt one-hour action

to initiate unit shutdown for conditions that could prevent the

injection of a CLA into the core. For boron concentration outside

limits, a 72-hour action to restore CLA concentration is allowed

because the CLA can still perform the core injection safety

function. The removal of surveillance requirements (SRs) for

verifying automatic opening features for the CLA isolation valves

does not impact the required TS alignment that is assumed in the

safety analysis. The instrumentation calibration and functional test

SRs have also been removed based on the instrumentation only

providing CLA level and pressure indications for TS compliance and

not performing an accident mitigation function. The above changes do

not alter the required limits for CLA operability or system

configurations. These changes are consistent with NUREG-1431 and

provide acceptable flexability[sic] for CLA operability verification

and surveillance testing and reasonable actions for CLA

inoperability. Since no changes have been proposed that would change

the conditions assumed for the CLAs in the accident analysis, the

consequences of an accident will not be increased. The CLAs perform

accident mitigation functions and are not considered to be the

source of an accident. Therefore, since the plant configurations and

functions are unchanged by the proposed changes, the probability of

an accident will not be increased.

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

from any previously analyzed.

The proposed changes clarify existing CLA operability

requirements, modify action times for CLA inoperability, enhance and

simplify SRs, and remove surveillances that are not required to

verify the CLA's ability to perform safety functions. None of these

changes affect the operation of the plant or the CLA configuration

and accident mitigation capabilities. Therefore, since the CLAs will

continue to support the plant as before, these proposed changes will

not create a new or different kind of accident.

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

The CLA requirements for volume, pressure, boron, and isolation

valve position are not changed by the proposed request. The CLAs

will continue to provide the same safety function capabilities as

assumed in the safety analysis. Therefore, no reduction in the

margin of safety will result from these chanes because CLA functions

are unchanged.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: September 9, 1994 (TS 94-08)

Description of amendment request: The proposed change would add

``main steam vaults'' to the footnote of Surveillance Requirement

4.6.1.1. This would allow inspection of the valves, blind flanges, and

deactivated automatic valves located in the vaults that are required to

be in the closed position during accident conditions and that are

locked, sealed, or otherwise secured in the closed position, on a cold

shutdown frequency rather than every 31 days.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determine that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.93(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change will exempt containment isolation valves in

the east and west main steam valve vaults from examination every

thirty one days if those valves are locked, sealed or otherwise

secured. The valves and flanges that are located inside the main

steam valve vaults and are required to be closed during accident

conditions, will be verified in their required position during cold

shutdown and will be secured in this position. The environmental

conditions in these areas ensure they will be low traffic areas

where the probability of misalignment or manipulation is remote.

Loss of containment integrity is not considered to be an initiator

of any accident. This change does not affect any accident analysis

assumptions or results for SQN. Therefore, there is no increase in

the probability or consequences of an accident previously evaluated,

as a result of this change.

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

from any previously analyzed.

This revision will not change any plant equipment, system

configurations, or accident assumptions. The appropriate components

in the valve vaults will continue to be verified in the closed

position and locked, sealed, or otherwise secured. The physical

congestion and high temperatures in the area will be effective in

maintaining this as a low traffic area that will contribute to the

low probability of misalignment or manipulation of these components

between inspections. Therefore, this change will not affect the

safety function of these components and will not create the

possibility of a new or different kind of accident.

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

The proposed change is consistent with current SQN accident

analysis assumptions since only the time interval between

performances of the surveillance is being extended. This change will

not impact any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: July 14, 1994

Description of amendment request: The proposed changes to the

Technical Specifications would remove the remaining references to

cycle-specific parameters in Technical Specification 3.12.A.2 and

associated Technical Specification Figures 3.12-1A and 1B. These

figures and the control bank insertion limits are presently specified

in the Core Operating Limits Report (COLR). The NRC-approved

methodologies presently listed in the Technical Specifications are used

to calculate and evaluate the parameter limits presented in the COLR

for each reload core.

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:

Specifically, operation of Surry Power Station in accordance

with the Technical Specification changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. The removal of the

remaining reference to cycle-specific core operating limits and

Technical Specification Figures 3.12-1A and 1B, from the Surry

Technical Specifications has no influence or impact on the

probability or consequences of any accident previously evaluated.

The proposed amendment is administrative in nature in that it

corrects omissions from a previously approved amendment. This change

has no impact on actions to be taken when or if limits are exceeded

as is required by the current Technical Specifications. Each

accident analysis addressed in the Surry UFSAR [Updated Final Safety

Analysis Report] will be examined with respect to changes in cycle-

dependent parameters, which are determined by application of NRC-

approved reload design methodologies. The impact of these parameter

changes on transient results is then evaluated to ensure that the

results remain bounded by respective transient analysis acceptance

criteria. This examination, which is performed per the requirements

of 10 CFR 50.59, ensures that future reloads will not involve an

increase in the probability or consequences of an accident

previously evaluated.

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

from any accident previously evaluated. The removal of the remaining

reference to cycle-specific core operating limits and Technical

Specification Figures 3.12-1A and 1B has no influence or impact, nor

does it contribute in any way to the probability or consequences of

any accident previously evaluated. No safety-related equipment,

safety function, or plant operating characteristic will be altered

as a result of the proposed changes. This cycle-specific variable

(control bank insertion limits) is calculated using NRC approved

methods, and the results are submitted to the NRC for information in

accordance with Technical Specification 6.2. The Technical

Specifications will continue to require operation within the

required core operating limits, and appropriate actions will be

taken when or if any of these limits are exceeded. Therefore, the

proposed amendment does not in any way create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

margin of safety is not affected by this administrative change which

removes the remaining reference to cycle-specific core operating

limits and Technical Specification Figures 3.12-1A and 1B from the

Technical Specifications. The margin of safety presently provided by

current Technical Specifications remains unchanged. Appropriate

measures exist to control the values of these cycle-specific limits.

The proposed amendment continues to require operation within the

core limits which were developed from the NRC-approved reload design

methodologies. Further, the actions to be taken when or if limits

are violated remain unchanged. Development of limits for future

reloads will continue to conform to those methods described in NRC-

approved documentation. In addition, each reload requires a 10 CFR

50.59 safety review to assure that operation of the unit within the

cycle-specific limits will not involve a reduction in any margin of

safety. Therefore, the proposed changes are administrative in nature

and do not impact the operation of Surry in a manner that involves 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: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

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

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

Virginia 23219.

NRC Project Director: Victor M. McCree (Acting)

Virginia Electric and Power Company, Docket Nos. 50-280, 50-281,

50-338,50-339, Surry Power Station, Units No. 1 and No. 2 Surry

County,Virginia and North Anna Power Station, Units No. 1 and No.

2, LouisaCounty, Virginia

Date of amendment request: September 6, 1994

Description of amendment request: The proposed changes would revise

the Technical Specifications (TS) for Surry 1&2 and North Anna 1&2.

Specifically, the proposed changes would revise the: (1) Management

Safety Review Committee (MSRC) review responsibilities regarding safety

evaluations and Station Nuclear Safety and Operating Committee (SNSOC)

meeting minutes and reports, and (2) SNSOC review responsibilities for

procedure changes. However, the changes now also state that the MSRC

will review safety evaluations, and the SNSOC will review procedure

changes, as programmatically discussed in the Updated Final Safety

Analysis Report (UFSAR).

The licensee's proposed changes revise and supersede the licensee's

original proposed changes dated December 27, 1993 and noticed in the

Federal Register on February 16, 1994, (59 FR 7700) for NA-1&2, and

March 16, 1994 (59 FR 12371) for Surry 1 & 2.

The North Anna and Surry Power Station Technical Specifications

presently address the organization and responsibilities of both the

onsite and offsite review groups, the SNSOC and the MSRC, respectively.

The responsibilities of the SNSOC include the review of new procedures

and changes to procedures that affect nuclear safety. The MSRC review

responsibilities include the review of safety evaluations and SNSOC

meeting minutes and reports. It is proposed that the extent of these

review activities be revised in the Technical Specifications to ensure

the two review groups are focusing on nuclear safety issues and not

spending an unnecessary amount of time on administrative activities of

minimal safety significance.

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:

[Specifically, operation in accordance with the proposed

Technical Specifications changes] will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. As administrative

changes, the proposed Technical Specifications changes have no

direct or indirect effect on accident precursors. No plant

modifications are being implemented and operation of the plant is

unchanged. SNSOC review of new procedures and procedure changes that

require a safety evaluation ensures that activities that could

affect nuclear safety are being properly reviewed. The MSRC's

overview of representative samples of safety evaluations and SNSOC

meeting minutes and reports based on performance ensures these

programs are being properly implemented and nuclear safety is not

being compromised; or

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

from any accident previously evaluated since physical modifications

are not involved and systems and components will be operated as

before the change. The proposed changes are wholly administrative in

nature and have no impact on plant operations or accident

considerations. These changes modify the scope of SNSOC review of

procedure changes and MSRC's review functions concerning safety

evaluations and SNSOC meeting minutes and reports. Procedure changes

will continue to receive management review in accordance with

administrative procedures, however, only changes that require a

safety evaluation will require SNSOC approval. MSRC review of

representatives samples of safety evaluations and SNSOC meeting

minutes and reports based on performance will continue to provide

adequate assurance that nuclear safety is being properly considered;

or

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

defined in the basis of any Technical Specification since the

responsibilities of the SNSOC and MSRC are not addressed by the

existing Technical Specification Bases, nor are review requirements

for procedures. The proposed changes are administrative in nature

and have no impact on, nor were they considered in, existing UFSAR

accident analyses. Safety significant procedure changes, i.e.,

changes that require a safety evaluation to be prepared, will

continue to be reviewed by SNSOC, as will new procedures. Procedure

changes still require cognizant management approval and preparation

of an activity screening to determine whether or not the change

impacts nuclear safety. This ensures activities important to nuclear

safety are being appropriately reviewed. The effectiveness of the

safety evaluation program, and the thoroughness of SNSOC meetings

and reports will be assured through the MSRC's plant overview

function which is based on observed performance.

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 locations: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185, and The Alderman

Library, Special Collections Department, University of Virginia,

Charlottesville, Virginia 22903-2498.

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

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

Virginia 23219.

NRC Project Director: Victor McCree, Acting

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: August 24, 1994

Description of amendment request: The proposed amendment would

revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)

3.1.b.1 and Figure TS 3.1-4 regarding Low Temperature Overpressure

(LTOP) protection for the reactor coolant pressure boundary. Currently,

the TS specify the LTOP requirements through the end of operating cycle

20 or 17.14 effective full power years. The proposed change extends the

LTOP requirements through the end of operating cycle 21 or 18.40

effective full power years. The Basis Section would also be modified to

reflect these 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 change was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed change will not:

1) involve a significant increase in the probability or

consequences of an accident previously evaluated.

The use of RG 1.99, Revision 2, Regulatory Position C.2 does not

modify the reactor coolant system pressure boundary, nor make any

physical changes to the facility design, material, construction

standards, or setpoints. The probability of a LTOP event occurring

is independent of the pressure temperature limits for the RCS

pressure boundary. Therefore, the probability of a LTOP event

occurring remains unchanged.

The use of predicted fluence values through the end of operating

cycle 21 is appropriately considered within the calculations in

accordance with standard industry methodology previously docketed

under WCAP 13227. Revised flux values were used for Cycles 16, 17,

18 and 19 based on actual core reload designs. Previous cycles flux

values are the same as reported in WCAP 12333.

The calculation of pressure temperature limits in accordance

with approved regulatory methods provides assurance that reactor

pressure vessel fracture toughness requirements are met and the

integrity of the RCS pressure boundary is maintained. Similar

methodology was used in calculations to support approved amendment

108 to the Kewaunee Technical Specifications dated April 7, 1994.

The use of Regulatory Position C.2 and fluence values through

EOC 21 meet previously established criteria for protection of the

health and safety of the public. The consequences of a LTOP

transient therefore, remain unchanged.

2) create the possibility of a new or different type of accident

from an accident previously evaluated.

The use of Regulatory Position C.2 and fluence through EOC 21

does not modify the reactor coolant system pressure boundary, nor

make any physical changes to the LTOP setpoint or system design.

Therefore, no new failure mechanisms are created that could

create the possibility of an accident of a new or different type.

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

The Appendix G pressure temperature limitations are calculated

in accordance with regulatory requirements and calculational

limitations specified in RG 1.99, Revision 2. RG 1.99, Revision 2,

is an acceptable method for implementing the requirements of 10 CFR

50 Appendices G and H. Similar methodology was used in calculations

to support approved amendment 108 dated April 7, 1994. The reactor

coolant pump starting restrictions of TS 3.1.a.1.c remain in place.

The revised calculations meet the NRC acceptance criteria for

the LTOP setpoint and system design as described in NRC Safety

Evaluation Report (SER) dated September 6, 1985 which concluded that

``the spectrum of postulated pressure transients would be

mitigated...such that the temperature pressure limits of Appendix G

to 10 CFR 50 are maintained.''

The use of Regulatory Position C.2, meets previously established

criteria for the pressure temperature limits for the LTOP system and

setpoint. Thus, the margin of safety as described in the NRC SER is

not reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Wisconsin

Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin

54301.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P. O. Box 1497, Madison, Wisconsin 53701-1497.

NRC Project Director: John N. Hannon

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: September 7, 1994

Description of amendment request: The proposed amendment would

revise Kewaunee Nuclear Power Plant (KNPP) Technical Specifications

(TS) by adding two new sections, TS Section 3.0 and TS Section 4.0,

with associated bases. TS Section 3.0 would establish the general

requirements applicable to each of the Limiting Conditions for

Operation (LCOs) within Section 3 of the KNPP TS. TS Section 4.0 would

establish the general requirements applicable to Surveillance

Requirements. The new requirements of TS 4.0.b would also affect TS

Sections 4.5, 4.6, 4.7, and Tables TS 4.1-2 and 4.1-3. The proposed TS

amendment incorporates guidance statements similar to Section 3.0/4.0

of NUREG-0452, ``Standard Technical Specifications for Westinghouse

Pressurized Water Reactors.''

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 were reviewed in accordance with the

provision of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will not:

1) involve a significant increase in the probability or

consequences of an accident previously evaluated.

The likelihood that an accident will occur is neither increased

or decreased by these TS changes. These TS changes will not impact

the function or method of operation of plant equipment. Thus, there

is not a significant increase in the probability of a previously

analyzed accident due to these changes. No systems, equipment, or

components are affected by the proposed changes. Thus, the

consequences of the malfunction of equipment important to safety

previously evaluated in the Updated Safety Analysis Report (USAR)

are not increased by these changes.

The proposed changes have no impact on accident initiators or

plant equipment, and thus, do not affect the probabilities or

consequences of an accident.

These changes are consistent with the requirements established

in the Westinghouse STS. Therefore, the proposed changes will not

significantly increase the probability or consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

The proposed TS changes would not create the possibility of a

new or different kind of accident from any accident previously

evaluated. The proposed changes do not involve changes to the

physical plant or operations. Since these changes do not contribute

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

or produce a new type of equipment malfunction. Also, these changes

do not alter any existing accident scenarios; they do not affect

equipment or its operation, and thus, do not create the possibility

of a new or different kind of accident.

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

Operation of the facility in accordance with the proposed TS

would not involve a significant reduction in a margin of safety. The

proposed changes do not affect plant equipment or operation. Safety

limits and limiting safety system settings are not affected by these

proposed changes. These changes are consistent with the Westinghouse

STS.

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 Wisconsin

Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin

54301.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P. O. Box 1497, Madison, Wisconsin 53701-1497.

NRC Project Director: John N. Hannon

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

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

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County,North Carolina

Date of amendments request: September 9, 1994 Brief description of

amendments request: The amendments change the Technical Specifications

to revise the frequency for verifying the position of the drywell-

suppression chamber vacuum breakers when the position indication is not

operable from at least once every 72 hours to at least once every 14

days.Date of publication of individual notice in Federal Register:

September 16, 1994 (59 FR 47648)

Expiration date of individual notice: October 3, 1994

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

Gulf States Utilities Company, Cajun Electric Power Cooperative,

and Entergy Operations, Inc., Docket No. 50-458, River Bend

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: September 8, 1994

Brief description of amendment: The proposed amendment would modify

Technical Specification 3.10.2, to permit the bypassing of the rod

withdrawal limiter notch constraints while performing fuel power

suppression testing. This modification to the technical specification

will allow River Bend Station to search for and identify the location

of leaking fuel bundles, during power operating conditions, so that

appropriate actions can be taken to prevent further degradation.

Date of publication of individual notice in Federal Register:

September 16, 1994 (59 FR 47652)

Expiration date of individual notice: October 17, 1994

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

Gulf States Utilities Company, Cajun Electric Power Cooperative,

and Entergy Operations, Inc., Docket No. 50-458, River Bend

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: September 12, 1994

Brief description of amendment: The proposed amendment would revise

the formula for calculating the average power range monitor (APRM) flow

biased simulated thermal power-high reactor trip and flow biased

neutron flux-upscale control rod block trip setpoints T-factor

specified in Technical Specification (TS) 3/4.2.2. The proposed changes

are necessary to support implementation of recommendations contained in

NRC Generic Letter 94-02, ``Long-Term Solutions and Upgrade of Interim

Operating Recommendations for Thermal-Hydraulic Instabilities in

Boiling Water Reactors.''

Date of publication of individual notice in Federal Register:

September 21, 1994 (59 FR 48456)

Expiration date of individual notice: October 21, 1994

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

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: September 9, 1994

Description of amendment request: The proposed amendment would

revise the Technical Specifications to modify surveillance requirements

by increasing the acceptance criterion for the closure of the main

steam isolation valves from 5 seconds to 10 seconds.

Date of publication of individual notice in Federal Register:

September 19, 1994 (59 FR 47960).

Expiration date of individual notice: October 19, 1994

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

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

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

Date of application for amendments: September 8, 1994 (TS 94-14)

Brief description of amendments: The amendment would separate the

portion of the steam generator tubing from the end of the tube up to

the start of the tube-to-tubesheet weld from the remainder of the tube

for the purposes of sample selection and repair when defects are found

in this section of a steam generator tube.

Date of publication of individual notice in the Federal

Register:September 19, 1994 (59 FR 47962)

Expiration date of individual notice: October 19, 1994

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennesee 37402.

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

Street, NW., Washington, DC 20555, and at the local public document

rooms for the particular facilities involved.

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

50-529 and STN 50-530, Palo Verde Nuclear Generating Station, Unit

Nos. 1, 2 and 3, Maricopa County, Arizona

Date of application for amendments: August 5, 1993

Brief description of amendments: The amendments change the phrase

``Pressurizer Pressure - Wide Range'' to ``Reactor Coolant System

Pressure - Wide Range'' in item 4 of TS Table 3.3-10 and item 4 of

Table 4.3-7. These amendments will clarify the instrumentation required

and eliminate potential confusion between the reactor coolant system

pressure instruments and the pressurizer pressure instruments.

Date of issuance: September 21, 1994

Effective date: September 21, 1994

Amendment Nos.: 81, 68, and 53

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74:

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 29, 1993 (58

FR 50962) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 21, 1994. No

significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

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

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

County, Maryland

Date of application for amendments: June 8, 1994

Brief description of amendments: The amendments revise Technical

Specification Section 4.7.1.2.c to extend the interval for three

Auxiliary Feedwater surveillance requirements from 18 to 24 months.

Date of issuance: September 26, 1994

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

30 days.

Amendment Nos.: 197 and 174

Facility Operating License No. DPR-53 and DPR-69: Amendment revised

the Technical Specifications.

Date of initial notice in Federal Register: August 17, 1994 (59 FR

42334) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated September 26, 1994.No

significant hazards consideration comments received: No

Local Public Document Room location: Calvert County Library,

Prince Frederick, Maryland 20678.

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

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

County, Maryland

Date of application for amendments: November 2, 1993, as

supplemented on June 22, 1994

Brief description of amendments: The amendments revise the

Technical Specifications regarding surveillance requirements associated

with the emergency diesel generators (EDGs) which include the

following: 1) the surveillance interval is extended from 18 months to

24 months which is the current refueling cycle; 2) removes the

requirement to verify the EDGs speed; 3) exempts sequencer testing in

Modes 5 and 6; 4) deletes the reference to the specific 2000 hour

rating of the EDGs; and 5) allows the EDGs to be prelubricated prior to

being started in accordance with the vendors recommendation.

Date of issuance: September 27, 1994

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

30 days.

Amendment Nos.: 198 and 175

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 8, 1993 (58 FR

64599) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated September 27, 1994.No

significant hazards consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

PointNuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: April 29, 1994

Brief description of amendment: The amendment revises surveillance

intervals associated with initiation of auxiliary feedwater on steam

generator water level (low-low) and on trip of the main feedwater

pumps. These revisions are being made in accordance with the guidance

provided by Generic Letter 91-04, ``Changes in Technical Specification

Surveillance Intervals to Accommodate a 24-Month Fuel Cycle.''

Date of issuance: September 23, 1994

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

30 days.

Amendment No.: 175

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 17, 1994 (59 FR

42335) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 23, 1994.No

significant hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

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

Date of application for amendment: December 10, 1993, as

supplemented by letter dated August 11, 1994.

Brief description of amendment: The amendment revises Technical

Specification (TS) Section 5.3.A., ``Reactor Core,'' to allow the use

of VANTAGE + fuel with ZIRLO cladding and of fuel with filler rods to

permit fuel reconstitution. The amendment also revises the Basis for TS

Section 2.1, ``Safety Limit: Reactor Core,'' to more accurately

describe the basis of the departure from nucleate boiling correlations

and how they are applied to ensure that the design criteria are met.

Date of issuance: September 29, 1994

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

within 30 days.

Amendment No.: 176

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 2, 1994 (59 FR

10003) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 29, 1994.No

significant hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling

Water Reactor, La Crosse, Wisconsin

Date of application for amendment: November 5, 1993 (Reference LAC-

13320) as supplemented August 3, 1994, (Reference LAC-13420).

Brief description of amendment: This amendment modified the

Technical Specifications (TS) incorporated in Possession-Only License

No. DPR-45 in accordance with a revision of 10 CFR Part 20 (56 FR

23360). In addition, there were minor clerical changes to correct

oversights from previous amendments.

Date of issuance: September 27, 1994.

Effective date: This license amendment is effective as of the date

of its issuance and must be fully implemented no later than 30 days

from the date of issuance.

Amendment No.: 68.Possession-Only License No. DPR-9: The amendment

revised the TS.

Date of initial notice in Federal Register: January 5, 1994 (59 FR

618) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 27, 1994.No significant hazards

consideration comments received: No.

Local Public Document Room location: La Crosse Public Library, 800

Main Street, La Crosse, Wisconsin 54601.

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 application for amendments: November 11, 1993, as

supplemented February 23, April 12 and July 29, 1994.

Brief description of amendments: The amendments reflect the

consolidation of the Quality Verification Department with the Nuclear

Generation Department that realigned the Nuclear Safety Review Board to

report to the Senior Nuclear Officer, change a reference from Semi-

Annual to Annual, change an organizational unit term from ``group'' to

``division,'' modify titles of positions designated to approve

modifications and clarify the responsibilities of the Safety Assurance

Manager.

Date of issuance: September 23, 1994

Effective date: September 23, 1994

Amendment Nos.: 124 and 118

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

revised the Technical Specifications.

Date of initial notice in Federal Register: January 5, 1994 (59 FR

618) The February 23, April 12 and July 29, 1994 letters provided

clarifying information that did not change the scope of the November

11, 1993, application and the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated September 23,

1994.No significant hazards consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of application for amendments: November 11, 1993, and

supplemented February 23, April 12 and July 29, 1994.

Brief description of amendments: The amendments reflect the

consolidation of the Quality Verification Department with the Nuclear

Generation Department that realigned the Nuclear Safety Review Board to

report to the Senior Nuclear Officer, change a reference from Semi-

Annual to Annual, change an organizational unit term from ``group'' to

``division,'' modify titles of positions designated to approve

modifications and clarify the responsibilities of the Safety Assurance

Manager.

Date of issuance: September 22, 1994

Effective date: September 22, 1994

Amendment Nos.: 148 and 130

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 19, 1994 (59 FR

2865) The February 23, April 12 and July 29, 1994, letters provided

clarifying information that did not change the scope of the November

11, 1993, application and the initial proposed no significant hazards

considerationdetermination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated September 22,

1994. No significant hazards consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

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

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

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

Pennsylvania

Date of application for amendments: June 2, 1994

Brief description of amendments: These amendments revise the

Appendix A TSs relating to reactor coolant leakage and leakage

detection systems in an effort to bring TS sections 3/4.4.6.1 and 3/

4.4.6.2 closer to NRC's Improved Standard TSs. A new TS, Section 3/

4.5.5 for Unit 1 and 3/4.5.4 for Unit 2, is added to address Seal

Injection Flow.

Date of issuance: September 22, 1994

Effective date: September 22, 1994

Amendment Nos.: 183 and 64

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

revised the Technical Specifications.

Date of initial notice in Federal Register: August 3, 1994 (59 FR

39585) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 22, 1994No significant

hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: February 9, 1993, as supplemented by

letter dated July 22, 1994.

Brief description of amendment: The amendment changed the Appendix

A Technical Specifications by revising Specifications 3.0.4, 4.0.3, and

4.0.4 in accordance with the intent of Generic Letter 87-09.

Date of issuance: September 20, 1994

Effective date: September 20, 1994

Amendment No.: 99

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

Technical Specifications.

Date of initial notice in Federal Register: August 17, 1994 (59 FR

42341) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 20, 1994.No

significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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 application for amendments: April 28, 1994, and

supplemented by letter dated July 29, 1994.

Brief description of amendments: 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 the following

NRC guidelines:A. Generic Letter (GL) 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements for

Testing During Power Operation.'' 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,''

Date of issuance: September 21, 1994

Effective date: September 21, 1994

Amendment Nos.: 75 and 54

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

revised the Technical Specifications.

Date of initial notice in Federal Register: August 3, 1994The

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

Safety Evaluation dated September 21, 1994.No significant hazards

consideration comments received: No

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830

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 application for amendments: June 24, 1994.

Brief description of amendments: The amendments revise the values

of Z and S in Technical Specification 2.2-1 for the Pressurizer

Pressure-Low and -High trip set-points (Table 2.2-1, Functional Units 9

and 10) to allow the use of Tobar, Veritrak, or Rosemount pressure

transmitters.

Date of issuance: September 22, 1994

Effective date: September 22, 1994

Amendment Nos.: 76 and 55

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

revised the Technical Specifications.

Date of initial notice in Federal Register: August 22, 1994 (59 FR

43143) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 22, 1994. No

significant hazards consideration comments received: No

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket No.

50-321, Edwin I. Hatch Nuclear Plant, Unit 1, Appling County,

Georgia

Date of application for amendment: August 16, 1994, as supplemented

September 20, 1994

Brief description of amendments: The amendment makes a one-time

change to Technical Specification (TS) 3.9.C for Hatch Unit 1 regarding

the emergency diesel generator (DG) operability requirements during

reactor shutdown conditions. Current TS 3.9.C requires that two DGs be

operable during reactor shutdown when a core or containment cooling

system is required to be operable. The amendment revises the current

requirement such that only one emergency DG is required to be aligned

to its associated core or containment cooling system during a specific

time of the outage. During this time period the decay heat removal

(DHR) system will be in service. The DHR system, which is completely

independent of the existing shutdown cooling system, is powered by the

Baxley substation and has its own DG as a b

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