Biweekly Notice

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving

No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from January 22, 1994, through February 4, 1994.

The last biweekly notice was published on February 2, 1994 (59 FR

4933).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 P-223, Phillips Building, 7920 Norfolk Avenue,

Bethesda, 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 March 18, 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.

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

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

Date of amendment request: January 24, 1994

Description of amendment request: The proposed amendment would

implement Line Item 5.9 of NRC Generic Letter 93-05, ``Line Item

Technical Specification Improvements to Reduce Surveillance

Requirements for Testing During Power Operation,'' which recommends

licensees consider deleting the requirements to perform response time

testing for selected instrumentation in the isolation system where the

required time corresponds to the diesel start time.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

LaSalle has evaluated the proposed Technical Specification

Amendment. Based upon the criteria for defining a Significant

Hazards Consideration established in 10 CFR 50.92(c), operation of

LaSalle County Station in accordance with the proposed amendment

will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The proposal seeks to eliminate response time testing

requirements for selected instrumentation in the isolation system.

The proposal does not introduce changes in the response times

themselves. The probability and consequences of an accident

previously evaluated are not increased because accepted licensing

criteria are maintained. The requirements for channel checks,

functional tests, calibrations, and logic system functional tests

are not altered by this proposal. The ability to detect degrading

trends of response times is available via the above Technical

Specification required tests. Therefore, the response times of these

systems will be maintained within the acceptance limits assumed in

plant safety analyses and required for successful mitigation of an

initiating event because of the continued Technical Specification

testing.

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

from any accident previously evaluated because:

The proposal does not change component or system interactions.

Accident analyses assume a loss of AC power which is restored by

startup of emergency diesel generators. The 13 second interval

associated with the restoration of AC power, which establishes the

response time for the isolation functions, is maintained. The

starting, sequencing, and loading functions associated with the

diesel generators is not affected by the proposed change. The

response times include the instrument response times, which are

typically measured in fractions of a second, and the response times

of the actuation logic circuits, which are typically less than a

second. These times are small in comparison to the diesel generator

start time (13 seconds). The ability of the isolation system to

perform its intended function to mitigate the consequences of an

initiating event within the acceptance limits assumed in plant

safety analyses is not altered by the proposed change.

3) Involve a significant reduction in a margin of safety

because:

The proposal does not involve the relaxation of any criteria

identified in the SAR or reduce any of the requirements of Technical

Specifications. The proposed revision does not affect licensing

acceptance limits associated with accidents. With the exception of

MSIVs, the safety analyses do not address individual sensor response

times or the response times of the logic systems to which the

sensors are connected. These analyses conservatively establish the

margin of safety. Deleting the requirement to perform unnecessary

response time testing does not affect the results of accident and

transient analyses. Plant and system response to an initiating event

will remain in compliance within the assumptions of safety analyses.

The proposed change does not increase the probability or

consequences of an accident, and there is no impact on equipment

important to safety or systems, structures or components. There is

no associated change to the type, amount, or control of radioactive

effluents, nor is there an associated increase in individual or

cumulative occupational radiation exposure. There is no effect upon

the capabilities of the associated systems to perform their intended

functions within the allowed response times assumed in safety

analyses. Therefore, the margin of safety is preserved.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Public Library of Illinois

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

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: James E. Dyer

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County, Connecticut

Date of amendment request: January 17, 1994

Description of amendment request: Connecticut Yankee Atomic Power

Company (CYAPCO) proposes to remove Technical Specification 3/4.4.12,

``Failed Fuel Rods'' and its associated BASES Section 3/4.4.12.

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 (SHC), which is presented below:

The proposed changes do not involve an SHC consideration because

the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

A review of the accidents detailed in the Updated Final Safety

Analysis Report, Chapter 15, was undertaken to determine if they

were impacted by the proposed change. The review indicated that the

previously evaluated accidents were not impacted by the proposed

license amendment.

All fuel design and performance criteria are the same for Cycle

18 as in previous cycles. All criteria will continue to be met and

no new single-failure mechanisms will be created. This change does

not involve any alterations to plant equipment or procedures which

would affect any operational modes or accident assumptions. This

proposed license amendment does delete a technical specification

that is no longer considered necessary. This deletion is prompted by

the replacement of stainless steel clad fuel with zircaloy clad

fuel. The zircaloy clad fuel, if it experiences damage, will release

iodine into the primary system. Any iodine released is covered

within the guidelines specified in the existing Technical

Specification 3/4.4.8, ``Specific Activity.'' This specification

will ensure that operation does not continue with radiochemistry

values that exceed those assumed in our accident assumptions. The

existing Technical Specification of specific activity along with the

zircaloy clad fuel will ensure that a significant increase in the

probability or consequences of an accident previously evaluated is

not present.

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

from any previously analyzed.

The possibility of an accident or malfunction of a different

type than any evaluated previously in the UFSAR [Updated Final

Safety Analysis Report] is not created. Since there are no changes

in the way the plant is operated, the potential for an unanalyzed

accident is not created. No new failure modes are introduced.

The presence of defective fuel rods and the resultant iodine

release would only affect potential offsite doses. This proposed

license amendment does not increase the radiochemistry limits, but

does revert the technical specifications back to the standard

methodology and limitations that were unable to be used because of

the stainless steel clad fuel. These new limitations will continue

to ensure that doses remain within the limits prescribed.

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

The proposed changes do not have any adverse impact on the

protective boundaries. The margin of safety, as defined in the basis

for any technical specification, is not reduced. The proposed

changes do not adversely impact any of the safety systems, nor do

they increase the number of challenges to the safety systems.

The limit of 160 defective rods was chosen to be consistent with

initial conditions assumed for the radiological design basis. The

elimination of this specification is acceptable since the basis for

the initial condition can be supported by the use of zircaloy clad

fuel as opposed to the unique stainless steel clad. If future fuel

defects are debris induced, the dose equivalent iodine will be

within expected radiochemistry values and the resulting doses will

be bounded. Therefore, there is no reduction in the margin of safety

as defined in the basis of any technical specification with the

deletion of the defective fuel rod technical specification.

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: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457.

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

Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-

3499.

NRC Project Director: John F. Stolz

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

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

Date of amendment request: December 6, 1993

Description of amendment request: The proposed amendment request

would revise the Technical Specifications (TSs) to provide a temporary

one-time revision to the Definition Section of the TS. Specifically, a

footnote is added in the Definition Section of the TS which is

applicable to TS 1.2.1, ``Cold Shutdown Condition,'' changing Tavq

less than or equal to 200 deg.F to less than or equal to 250 deg.F and

TS 1.2.2, ``Hot Shutdown Condition,'' changing Tavq greater than

200 deg.F to greater than 250 deg.F. The footnote further states that

the change is for the one time, fuel out, chemical decontamination

program. This program is currently scheduled for the upcoming 1995

refueling outage of the Indian Point Nuclear Generating Unit 2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

Approval of the proposed one time change to the Technical

Specification definition of cold shutdown for purposes of performing

the full RCS [reactor coolant system] chemical decontamination

without fuel in the reactor would provide relief from unnecessary

technical specification action statements that are based on fuel in

the reactor. Credible accidents with significant consequences are

practically eliminated with the removal of the reactor fuel during

the performance of the FSD [full reactor coolant system chemical

decontamination]. In addition, specific actions would be taken in

accordance with the requirements of the NRC approved WCAP-12932-A

Rev. 2 to ensure that RCS and affected interfacing systems integrity

are preserved. Thus, system capability within established accident

scenarios would not be compromised. The proposed amendment would

therefore not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

any previously analyzed has not been created.

As noted above, the proposed amendment seeks to eliminate

unnecessary Technical Specification action requirements during the

performance of full RCS chemical decontamination. These actions are

unnecessary because there will be no fuel in the reactor and the RCS

and other affected systems will be operated under conditions well

within their design capability during the implementation of this

process. In addition, the FSD effort will be conducted in accordance

with the requirement(s) of the NRC approved Westinghouse topical

report WCAP-12932-A Rev. 2. Accidents involving failures of the

decontamination process system will not exceed the bounding

conditions for any previously established accidents involving

failure of a radwaste system. Accordingly, the possibility of a new

or different kind of accident from any previously analyzed will not

be created.

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

The proposed amendment provides relief from technical

specification actions in the performance of the FSD which become

unnecessary when there is no fuel in the reactor. The change will

not adversely impact any Technical Specification required systems,

structures or components. The design capability of systems,

structures or components impacted will not be reduced. Consequently,

no significant reduction in the margin of safety for any system,

structure, or component is involved.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

New York, New York 10003.

NRC Project Director: 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: January 18, 1994

Description of amendment request: The proposed amendments remove

the tables of containment penetration conductor overcurrent protective

devices from the Technical Specifications (TS) in accordance with the

guidance contained in Generic Letter 91-08, ``Removal of Component

Lists from Technical Specifications.'' The tables would be relocated to

Chapter 16 of the Catawba Final Safety Analysis Report (Selected

Licensee Commitments Manual). In addition, the licensee proposes the

removal of an obsolete footnote to TS 4.8.4. The footnote, which made

TS 4.8.4.a initially effective following the first refueling outage of

Unit 1, is no longer needed.

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. Relocating the component lists of containment penetration

conductor overcurrent protective devices from the technical

specifications to the [Selected Licensee Commitments] SLC Manual

(with all attendant required technical specification changes as

described previously and also including removal of the above

described obsolete footnote) has no impact upon either the

probability or consequences of any accident. No plant equipment is

affected by the proposed change. No equipment is being added or

deleted from the lists; only the source document for the lists is

being changed. Any future changes to the lists (i.e., changes to the

plant) will be subject to the provisions of 10CFR50.59 and also

subject to the change control provisions of Chapter 6 of Catawba's

Technical Specifications.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. No accident causal mechanisms are affected by the

proposed change, as no change to the plant is being proposed. In

addition, no change to the manner in which the plant is operated is

being made. Finally, no changes to plant procedures are being made

which would affect any accident causal mechanisms.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. The proposed change has no impact

upon any safety margin. The proposed change is consistent with the

guidance provided in Generic Letter 91-08 and the control provisions

utilized as a result of relocating the subject component lists are

at least as stringent as those set forth in the generic letter.

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: Loren R. Plisco, Acting

Duke Power Company, Docket No. 50-413, Catawba Nuclear Station,

Unit No. 1, York County, South Carolina

Date of amendment request: January 10, 1994

Description of amendment request: The proposed amendment would

revise Technical Specifications 2.0 and 3/4.2 which currently requires

the determination of the reactor coolant system flow rate by precision

heat balance measurement at least once per 18 months. Date of

publication of individual notice in Federal Register: January 26, 1994

(59 FR 3743)

Expiration date of individual notice: February 25, 1994

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 amendment request: January 10, 1994

Description of amendment request: The amendments would change the

method of measuring the reactor coolant system flow rate (Technical

Specifications 2.0 and 3/4.2) during the 18-month surveillance for

McGuire, Units 1 and 2.

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) This amendment will not significantly increase the probability

or consequence of any accident previously evaluated.

No component modification, system realignment, or change in

operating procedure will occur which could affect the probability of

any accident or transient. The change in method of flow measurement

will not change the probability of actuation of any Engineered

Safeguard Feature or other device. The actual flow rate will not

change. The consequences of previously-analyzed accidents will not

change as a result of the new method of flow measurement.

(2) This amendment will not create the possibility of any new or

different accidents not previously evaluated.

No component modification or system realignment will occur which

could create the possibility of a new event not previously

considered. The elbow taps are already in place, and are used to

monitor flow for the Reactor Protection System. They will not

initiate any new events.

(3) This amendment will not involve a significant reduction in a

margin of safety.

As described in [the licensee's application], the change in

method of RCS flow measurement will provide a more accurate

indication of the flow. The actual flow rate will not be affected.

The revised setpoints for low reactor coolant flow are driven by

changes to statistical allowances and do not represent substantive,

or less conservative, changes. There is no significant reduction in

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Astkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Loren R. Plisco, Acting

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 23, 1993

Description of amendment request: The proposed amendment would

revise the Technical Specification (TS) for the following four items in

accordance with the guidance in Generic Letter (GL) 93-05 ``Line Item

Technical Specifications Improvements To Reduce Surveillance

Requirements For Testing During Power Operation''.1) GL Item 5.14

Radiation Monitors will change the channel functional test from monthly

to quarterly.2) GL Item 6.1 Reactor Coolant System (RCS) Isolation

Valves will increase the time from 72 hours to 7 days for remaining in

cold shutdown without leak testing the RCS isolation valves.3) GL Item

6.6 Pressurizer Heaters will change the verification of capacity from

at least once per 92 days to each refueling outage and will change the

demonstration of the emergency power supply from at least once per 18

months to at each refueling outage.4) GL Item 9.1 Auxiliary Feedwater

Pump and System Testing will change the frequency of these pumps from

once per 31 days on a staggered basis to quarterly on a staggered

bases.

All of the above are compatible with Waterford 3 plant operating

experience and are consistent with NUREG-1366, ``Improvement To

Technical Specification Surveillance Requirements,'' December 1992 and

the licensing basis for Waterford 3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change to increase the radiation monitoring

instrumentation channel functional test from monthly to quarterly

will have no effect on design basis accidents. The findings in

NUREG-1366 determined that this change will increase the

availability of radiation monitors.

The proposed change to increase the 72 hour time for remaining

in cold shutdown without leak testing the RCS isolation valves to 7

days will not affect any design basis accidents. NUREG-1366 findings

have determined that extending this interval does not significantly

alter the associated risk. In addition, the current requirement has

a potential for causing problems resulting from a hurried recovery.

The proposed change to the pressurizer heater capacity test

interval from quarterly to each refueling interval will have no

affect on any design basis accidents. The TS requires at least 2

groups of pressurizer heaters each having a nominal capacity of 150

kW. Waterford 3 has 8 groups of pressurizer heaters; two

proportional groups of 150 kW each, and 6 backup groups of 200 kW

each. An evaluation of past operating experience has shown the

availability of at least 6 groups of pressurizer heaters with a

minimum of 150 kW each.

The proposed change to extend the testing interval for the EFW

[emergency feedwater] pumps will have no affect on any design basis

accidents. The pumps will continue to be tested quarterly to the

same standards applied to safety related pumps as defined by the

ASME [American Society of Mechanical Engineers] Section XI Code.

Satisfactory completion of testing in accordance with the Code is

accepted as verification that safety related pumps will be available

to perform their intended function.

The proposed changes identified above are supported by the

findings identified in NUREG-1366 and consistent with the guidance

provided in Generic Letter 93-05. These line-item improvements are

intended to improve plant safety, decrease equipment degradation,

and remove unnecessary burden on personnel resources by reducing the

amount of testing that the TS require during power operation.

Therefore, the proposed changes identified above will not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

The changes identified above only affect the frequency of

surveillance testing. There are no changes that will alter operation

of the plant or the manner in which it is operated. Therefore, the

proposed changes will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes identified herein extend testing frequency

in an effort to improve plant reliability and safety. The proposed

changes are consistent with the findings in NUREG-1366, guidance in

Generic Letter 93-05 and plant operating experience. As such, the

proposed changes will preserve the established margin of safety for

the affected specifications. Therefore, the proposed changes will

not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

L Street NW., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

Georgia Power Company, Oglethorpe Power Corporation, Municipal

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

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

Burke County, Georgia

Date of amendment request: November 19, 1993

Description of amendment request: The proposed change would

relocate the requirements of Technical Specification 3/4.3.4, Turbine

Overspeed Protection, to Section 16.3 of the Vogtle Electric Generating

Plant, Units 1 and 2, Final Safety Analysis Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed change involves the relocation of the TS

[Technical Specification] requirements for the turbine overspeed

protection system to the VEGP [Vogtle Electric Generating Plant]

FSAR [Final Safety Analysis Report]. The requirements that will

reside in the FSAR will continue to ensure that the probability of

turbine missile generation is maintained below NRC limits as defined

in NUREG-1048, Appendix U. Since the turbine overspeed protection

system will remain capable of protecting the turbine from excessive

overspeed, the proposed change will have no effect on the

consequences of an accident previously evaluated.

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

or different kind of accident than any previously evaluated. The

proposed change does not involve any change to the configuration or

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

have been defined for any plant system or component. In addition, no

new limiting failures have been identified as a result of the

proposed change. The requirements for the turbine overspeed

protection system that will reside in the FSAR will ensure that the

system remains capable of protecting the turbine from excessive

overspeed. Therefore, the proposed change will not create the

possibility of a new or different kind of accident than any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety. The proposed change would allow the

requirements for the turbine overspeed protection system to be

relocated to the FSAR on the basis that the turbine overspeed

protection system does not meet the criteria of the NRC Final Policy

Statement on Technical Specifications Improvements for Nuclear

Reactors. The requirements that will reside in the FSAR for the

turbine overspeed protection system will ensure that the system

remains capable of protecting the turbine from excessive overspeed.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

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

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia 30308

NRC Project Director: Loren R. Plisco, Acting

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket No. 50-499 South Texas Project, Unit 2, Matagorda County,

Texas

Date of amendment request: January 25, 1994

Description of amendment request: The licensee proposes to make a

one-time change to the technical specifications to add new Technical

Specifications 3/4.10.6 and 3/4.10.7 to the Special Test Exemptions

section. The new TS would allow the restart of Unit 2 with expired

calibrations on the core exit thermocouples (CET) and the reactor

coolant system (RCS) resistance temperature detectors (RTD). This

amendment will also add a new Technical Specification to allow the

ascension to 75 percent rated thermal power with an expired precision

heat balance reactor coolant flow measurement.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

in the probability or consequences of a previously evaluated

accident.

The proposed change will allow the restart of [STP] Unit 2 with

Core Exit Thermocouples and Reactor Coolant System Resistance

Temperature Detectors technically inoperable due to expired

calibrations. The calibrations of these instruments can only be

completed when the Unit reaches Normal Operating Pressure and Normal

Operating Temperature in Mode 3. Once the calibrations of these

instruments are completed, this one time change will expire and all

of the existing applicable Limiting Conditions for Operations will

become effective immediately. Since industry and South Texas Project

Electric Generating Station experience has shown that the failure

mechanism for these types of instrument is complete failure as

opposed to a gradual drift, and there will be calibration points to

compare RTD readings to actual RCS temperature as the RCS

temperature increases, it is reasonable to expect these CETs/RTDs

will function as they did before their calibrations expired. For

this reason, all applicable functions, including COMS, Thot ,

Tcold, and Tavg are expected to operate normally. Because

normal operation of the instruments is expected and the only reason

for the instruments being declared inoperable is their expired

calibrations, this change does not involve a significant increase in

the probability or consequence of an accident previously evaluated.

The proposed change will also allow the restart of Unit 2 with

the precision heat balance RCS flow measurement surveillance

expired. This surveillance is used to confirm the values indicated

by the RCS flow meters. These instruments are calibrated every 18

months and the RCS flow meters will be checked every 12 hours to

ensure adequate flow prior to the completion of the precision heat

balance RCS flow measurement. Since this surveillance is only used

to confirm the reading of calibrated instruments and does not

involve any changes to the design or function of the instruments,

this change does not involve a significant increase in the

probability or consequence of an accident previously evaluated.

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

or different kind of accident from any previously evaluated.

The operations of Unit 2 with the CETs and RCS RTDs technically

inoperable due to expired calibrations, until these calibrations can

be completed in Mode 3, does not affect the design bases of the CETs

and RCS RTDs or any of the accident evaluations involving these

instruments. Since industry and South Texas Project Electric

Generating Station experience indicates that the failure mechanism

for these types of instruments is not a gradual drift but complete

failure, the reasonable expectation is the CETs/RTDs will function

as they did prior to their calibrations expiring.

Additionally, the operation of Unit 2 with the precision heat

balance RCS flow measurement surveillance expired does not affect

the design bases of the RCS flow meters or any of the accident

evaluations involving these instruments. This surveillance is used

to confirm the values indicated by the RCS flow meters. These

instruments are calibrated every 18 months and the RCS flow meters

will be checked every 12 hours to ensure adequate flow prior to the

completion of the precision heat balance RCS flow measurement.

Because normal operation of all of these instruments is

expected, these changes do not create the possibility of a new or

different kind of accident from any previously evaluated.

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

in the margin of safety.

The RCS RTDs are auctioneered to prevent a failed high or low

instrument from adversely influencing the safety of the plant. This

feature is still operable and will, along with normal operator

activities, provide assurance that the margin of safety is not

reduced by this change. In addition, the change does not affect the

design bases, accident analysis, reliability or capability of the

CETs/RTDs to perform their intended safety functions. The RCS flow

meters will be checked every 12 hours to ensure adequate flow prior

to the completion of the precision heat balance RCS flow

measurement.

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

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

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

request for amendment involves no significant hazards

consideration.Local Public Document Location: Wharton County Junior

College, J.M. Hodges Learning Center, 911 Boling Highway, Wharton,

Texas 77488

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

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

NRC Project Director: Suzanne C. Black

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: September 28, 1993

Description of amendment request: The proposed amendment would

revise the Cooper Nuclear Station (CNS) Technical Specifications to

modify the licensee's organizational structure by removing the

positions of ``Site Manager'' and ``Senior Manager of Operations.'' The

functions presently given in CNS Technical Specifications for the Site

Manager position will be assumed by the Vice President - Nuclear.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Evaluation

The proposed change removing the positions of Site Manager and

Senior Manager of Operations from the Technical Specifications is

administrative in nature. The functions and responsibilities of the

previous position of Site Manager presently given in the plant

Technical Specifications will be performed by the Vice President -

Nuclear. Additionally, with the reorganization, the Senior Manager

of Operations position is eliminated and therefore, this position is

also being removed. The provision in the Technical Specifications

for automatic shifting of Plant Manager responsibilities to the

Senior Manager of Operations has also been removed. The shifting of

Plant Manager responsibilities (in writing) to one of the Managers

at CNS who is qualified for this position remains in the Technical

Specifications. The position removals and responsibility transfers

in the organization do not affect plant design or operation, nor do

they affect the way any systems, structures, or components are

operated or maintained. The individual filling the position ``Vice

President - Nuclear'' is qualified to perform the assigned tasks and

responsibilities. Restructuring of the sentence in specification

6.2.B.6, is purely an administrative change. Also, this proposed

change does not alter the conditions or assumptions in any of the

Updated Safety Analysis Report (USAR) accident analyses. Since the

USAR accident analyses remain bounding, the consequences previously

evaluated are not adversely affected by the proposed change.

Therefore, it can be concluded that the proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. Does the proposed License Amendment create the possibility of

a new or different kind of accident from any accident previously

evaluated?

Evaluation

The proposed Technical Specification revision removes all

references to the position title of the Site Manager. The

responsibilities of this position presently given in the Technical

Specifications are being incorporated and performed by the position

``Vice President - Nuclear.'' Additionally, with the reorganization,

the Senior Manager of Operations position is eliminated and

therefore, this position is also being removed. The shifting of

Plant Manager responsibilities (in writing) to one of the Managers

at CNS who is qualified for this position remains in the Technical

Specifications. All given management activities will continue to be

performed by qualified individuals. Restructuring of the sentence in

specification 6.2.B.6 is purely an administrative change. This

change does not affect the design or operation of any system,

structure, or component in the plant, and is considered to be an

administrative change. Accordingly, no new failure modes have been

defined for any plant system or component important to safety, nor

has any new limiting failure been identified as a result of the

proposed change. Therefore, this proposed change does not create the

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

previously evaluated.

3. Does the proposed amendment involve a significant reduction

in the margin of safety?

Evaluation

This proposed amendment involves a change to the Administrative

Controls Section of the CNS Technical Specifications; specifically,

removal of two positions referenced in the organizational structure.

The Site Manager position is being deleted and the responsibilities

of this position listed in the Technical Specifications are being

performed by the Vice President - Nuclear. Additionally, with the

reorganization, the Senior Manager of Operations position is

eliminated and therefore, this position and responsibilities are

also being removed. The shifting of Plant Manager responsibilities

(in writing) to one of the Managers at CNS who is qualified for this

position remains in the Technical Specifications. All given

management activities, as described in the Technical Specifications,

will continue to be performed by qualified individuals.

Restructuring of the sentence in specification 6.2.B.6, is purely an

administrative change. The proposed change does not adversely impact

the plant's ability to meet applicable regulatory requirements. The

proposed change does not alter any means of plant operation, nor

does the proposed change involve any physical alterations to the

plant and does not affect any plant safety parameters or setpoints.

Therefore, this proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Asuburn Public Library, 118

15th Street, Auburn, Nebraska 68305

Attorney for licensee: Mr. G. D. Watson, Nebraska Public Power

District, Post Office Box 499, Columbus, Nebraska 68602-0499

NRC Project Director: William D. Beckner

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: December 10, 1993

Description of amendment request: The proposed amendment would

revise the Cooper Nuclear Station (CNS) Technical Specifications

Sections 3/4.21 ``Environmental/Radiological Effluents,'' and 6.5,

``Station Reporting Requirements,'' to change the frequency of the

reporting period of the ``Semiannual Radioactive Materials Release

Report'' from semiannual to annual and to extend the reporting

frequency of the Annual Design Change Report from an annual submittal

to annually or along with the Updated Safety Analysis Report (USAR)

updates required by 10 CFR 50.71(e). These proposed changes are

intended to make the CNS Technical Specifications consistent with the

current provisions of 10 CFR 50.36(a) and 10 CFR 50.59(b),

respectively.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Evaluation

The proposed changes are administrative in nature and makes the

Cooper Nuclear Station (CNS) Technical Specifications (T/S)

consistent with amended regulations of 10CFR50.36(a), and 10CFR

50.59(b) by reducing the submittal frequency of certain reports to

the NRC. The proposed revisions do not involve any change to plant

design, plant operation, or configuration of any plant equipment

that is used to mitigate the consequences of an accident previously

evaluated. Also, the proposed changes do not alter the conditions or

assumptions in any of the Updated Safety Analysis Report (USAR)

accident analyses. Since the USAR accident analyses remain bounding,

the radiological consequences previously evaluated are not adversely

affected by the proposed changes. As administrative changes, all

defined terms on the affected pages have been capitalized.

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

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. Does the proposed License Amendment create the possibility of

a new or different kind of accident from any accident previously

evaluated?

Evaluation

The proposed changes are administrative in nature and makes the

CNS T/S consistent with amended regulations of 10CFR50.36(a), and

10CFR50.59(b) by reducing the submittal frequency of certain reports

to the NRC. The proposed revisions do not involve any change to

plant design, plant operation, or configuration of any plant

equipment that is used to mitigate the consequences of an accident

previously evaluated. Accordingly, no new failure modes have been

created for any plant system or component important to safety nor

has any new limiting failure been identified as a result of the

proposed changes. Also, there will be no change in the types or

increase in the amount of effluents released offsite. As

administrative changes, all defined terms on the affected pages have

been capitalized. Therefore, the proposed change does not create the

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

previously evaluated.

3. Does the proposed change create a significant reduction in

the margin of safety?

Evaluation

The proposed changes are administrative in nature and do not

adversely impact the plant's ability to meet applicable regulatory

requirements related to liquid or gaseous effluents, and solid waste

releases. The proposed changes do not alter any administrative

controls over radioactive effluents, nor do the proposed changes

involve any physical alterations to the plant with respect to

radioactive effluents. These changes do not affect the meaning,

application, and function of the T/S requirements. The proposed

change will reduce the administrative burden of NRC reporting

without reducing the protection for public health and safety. As

administrative changes, all defined terms on the affected pages have

been capitalized. Therefore, the proposed change does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Asuburn Public Library, 118

15th Street, Auburn, Nebraska 68305

Attorney for licensee: Mr. G. D. Watson, Nebraska Public Power

District, Post Office Box 499, Columbus, Nebraska 68602-0499

NRC Project Director: William D. Beckner

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

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

Date of amendment request: January 6, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Tables 3.2.7, 3.6.2a, 4.6.2a,

3.6.2b and 4.6.2b to delete the main steam line isolation and automatic

reactor shutdown (reactor scram) functions of the Main Steam Line

Radiation Monitor. Conforming changes would also be made to the Bases

of these TSs and to the Bases for TS 2.1.2. The licensee stated that

the proposed changes would be consistent with the NRC's Improved

Standard Technical Specifications, NUREG-1433, and with NRC-approved

(Safety Evaluation, dated May 15, 1991) Boiling Water Reactor Owners'

Group Licensing Topical Report NEDO-31400A, dated July 9, 1987.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the functions being removed do not contribute to avoidance

of any previously evaluated accidents. Further, the changes have

been shown to have an insignificant impact on overall reactivity

control failure frequency. This insignificant impact is offset by

the relatively large reduction in core damage frequency realized by

the implementation of these changes. Hence, the probability or

consequences of previously evaluated accidents are not significantly

increased due to this change. To the contrary, as stated in the

topical report [NEDO-31400A] the changes provide a net improvement

in overall plant safety.

The proposed amendment does not involve a physical or procedural

change to any structure, component or system that significantly

affects the probability or consequences of any accident or

malfunction of equipment important to safety previously evaluated in

the Final Safety Analysis Report (Updated). The proposed amendment

will involve a change to reactor protection and isolation actuation

systems circuitry that will remove the automatic reactor shutdown

and Main Steam Line Isolation Valve closure functions of the Main

Steam Line Radiation Monitor. However, the physical changes will not

affect the remaining scram or vessel isolation functions.

[***T]he methods, procedures and assumptions used to perform the

eneric analyses in NED0-31400A are bounding for the Nine Mile Point

Unit 1 with regard to input values. Niagara Mohawk has also provided

in the evaluation reasonable assurance that significantly increased

levels of radioactivity in the main steam lines will be controlled

expeditiously to limit both occupational and environmental

exposures. The Main Steam Line Radiation Monitor alarm setpoints

will be set at 1.5 times the normal full power background dose rate

and should any monitor exceed its alarm setpoint, the reactor

coolant will be sampled to determine activity levels and the

possible need for additional corrective actions.

The offgas radiation monitor is a more sensitive monitor than

the Main Steam Line Radiation Monitor because the nitrogen-16

source, dominating the radiation signal to the Main Steam Line

Radiation Monitor, has decayed by the time the radiation monitor can

be affected by any increased levels of activity. Therefore, setting

the offgas radiation monitor at 1.5 times the nitrogen-16 background

dose rate is not reasonable since setting the monitor that low can

lead to spurious activations of the alarm.

Nine Mile Point Unit 1's monitor configuration, as described in

the FSAR, detects the concentration of the offgas as it flows

through the pipe. Thus, the detector is sensitive to fluctuations in

condenser air inleakage, which can have an appreciable impact on the

monitor readings, especially at readings as low as 1.5 times the

normal full power background. Therefore, Niagara Mohawk proposes to

set the alarm at five (5) times the normal full power background,

which is still very conservative compared to the value allowed by

Technical Specification 3.6.15.c., which is set based on Nine Mile

Point Unit 1's Offsite Dose Calculation Manual.

Niagara Mohawk believes that a setting of five (5) times the

normal full power background is extremely conservative and is low

enough to ensure detection of even minor fuel performance changes.

Furthermore, if the monitor alarms at this setpoint of five times

the normal full power background, the offgas will immediately be

sampled and analyzed, followed by an analysis of a reactor coolant

sample.

Furthermore, the analyses in the Licensing Topical Report

demonstrate that removal of the automatic reactor scram and Main

Steam Line Isolation Valve closure functions of the Main Steam Line

Radiation Monitor does not change the conclusions in the Final

Safety Analysis Report (Updated) that the calculated radiological

release consequences of the bounding control rod drop accident will

not exceed the acceptable dose limits specified in 10CFR[Part]100.

Therefore, Niagara Mohawk concludes that 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 previously evaluated.

The function of a Main Steam Line Radiation Monitor trip is to

detect abnormal fission produce release and isolate the main steam

lines, thereby stopping the transport of fission products from the

reactor to the main condenser. The monitors do not perform a

prevention function for any kind of accident.

The main steam line high radiation scram and main steam line

isolation functions were originally intended to mitigate, not

prevent, an existing accident scenario. However, the functions being

removed do not contribute to avoidance or mitigation of any

previously evaluated accidents since no credit is taken for these

functions in any design basis event for terminating the initiating

event or assuring the radioactive release remains within accepted

limits. The existence of a Main Steam Line Radiation Monitor trip

does not prevent the occurrence of a fuel failure event or any other

type of event. Elimination of these functions will not introduce a

new or different accident scenario.

The proposed amendment represents a change to the physical

configuration of the plant in that some reactor protection system

circuits will be modified to eliminate the main steam line high

radiation scram and main steam line isolation signals. However,

these changes will not affect the remaining scram or vessel

isolation functions. In all other respects, plant design and

operation remain unchanged.

Therefore, Niagara Mohawk Power Corporation concludes that the

proposed amendment will not create the possibility of a new or

different kind of accident from any previously evaluated.

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

proposed amendment will not involve a significant reduction in a

margin of safety.

The proposed changes do not involve a significant reduction in a

margin of safety because, as shown in the topical report, the

changes represent an overall improvement in plant safety in that the

core damage frequency is reduced. Safe operation of the plant is

enhanced by elimination of the unnecessary scram and isolation of

the reactor vessel. With implementation of these changes, the

primary heat sink remains available, a large transient on the vessel

and safety-related actuations is avoided, and the Offgas System

remains available to control the pathway of a potential release.

Therefore, Niagara Mohawk concludes that 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: Robert A. Capra

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone

NuclearPower Station, Unit 1, New London County, Connecticut

Date of amendment request: January 14, 1994

Description of amendment request: The proposed amendment corrects

an editorial error. Specifically, the amendment changes the reference

in Limiting Condition for Operation (LCO) 3.4.D from ``3.3.A through

C'' to ``3.4.A, 3.4.B, and 3.4.C.'' The amendment also changes the

associated bases to clarify the LCO minimum solution concentration

requirement of 11 weight percent and updates the excerpt from 10 CFR

50.62 to reflect the current text of the regulation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with 10 CFR 50.92, NNECO [Northeast Nuclear Energy

Company] has reviewed the proposed change and has concluded that it

does not involve a significant hazards consideration (SHC). The

basis for this conclusion is that the three criteria of 10 CFR

50.92(c) are not compromised. The proposed change does not involve

an SHC because the change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change only modifies an incorrect reference in

Section 3.4.D of the Technical Specifications. In practice, if

Specification 3.4.A, 3.4.B, or 3.4.C cannot be met, an orderly

shutdown is initiated. As currently written, the failure to meet the

requirements of Section 3.3 would also initiate a shutdown in

accordance with Section 3.4.D. This is not the intent of Section

3.4.D since Section 3.3 already has specific shutdown requirements.

This proposed change will correct Section 3.4.D so that it limits

the conditions under which a plant shutdown must be initiated to the

LCOs of the standby liquid control system. Therefore, this proposed

change will not increase the probability or consequences of an

accident.

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

from any previously evaluated.

The proposed change corrects an incorrect section reference.

There is no change to the operation or design of the plant, nor is

there any change to the operability requirements of either section.

The proposed change properly identifies the conditions under which

the plant must be shutdown if an LCO is not met for the standby

liquid control system. In practice, if Specification 3.4.A, 3.4.B,

or 3.4.C cannot be met, an orderly shutdown is initiated. Since

there is no change in plant operation or design, there is no

possibility of a different kind of accident.

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

The proposed change does not modify the design or function of

the plant, nor does it reduce operability requirements of either

Section 3.3 or 3.4. The proposed change only corrects an incorrect

section reference by identifying the correct shutdown requirements

for the standby liquid control system. Since there is no change to

plant operation or design and the shutdown requirements are not

reduced, there is 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: Learning Resources Center,

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

London Turnpike, Norwich, Connecticut 06360.

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

Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-

3499.

NRC Project Director: John F. Stolz

Northeast Nuclear Energy Company (NNECO), Docket Nos. 50-245, 50-

336 and 50-423, Millstone Nuclear Power Station, Units 1, 2 and 3,

New London County, Connecticut

Date of amendment request: December 22, 1993

Description of amendment request: The proposed amendments would

change the Technical Specification (TS) as follows:

1. Change the title of the Nuclear Station Director to Senior Vice

President - Millstone Station.

2. Remove the requirement to provide a copy of Plant Operations

Review Committee (PORC) and Site Operations Review Committee (SORC)

meeting minutes to the Executive Vice President - Nuclear. The Senior

Vice President - Millstone Station is being proposed to replace the

Executive Vice President - Nuclear for receipt of PORC and SORC meeting

minutes.

3. Make editorial changes to the Millstone Unit No. 1 TS Index.

4. Correct a typographical error in Section 6.2.1.d of the

Millstone Unit No. 1 TS.

5. Correct a typographical error in Section 6.5.3.1.a of the

Millstone Unit No. 3 TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is provided below:

The proposed changes do not involve an SHC because the changes

do not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

No design basis accidents are affected by these proposed

changes. The proposed changes are administrative and editorial in

nature to reflect a recent reorganization, removal of the Executive

Vice President - Nuclear from receipt of PORC and SORC meeting

minutes, addition of the Senior Vice President - Millstone Station

to the receipt of PORC and SORC meeting minutes, and editorial

changes to the Millstone Unit Nos. 1 and 3 Technical Specifications.

No safety systems are adversely affected by the proposed changes,

and no failure modes are associated with the changes. Therefore,

there is no impact on the probability of occurrence or the

consequences of any design basis events.

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

from any previously evaluated.

Since there are no changes in the way the plant is operated, the

potential for an unanalyzed accident is not created. There is no

impact on plant response, and no new failure modes are introduced.

These proposed administrative and editorial changes have no impact

on safety limits or design basis accidents, and they have no

potential to create a new or unanalyzed event.

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

The changes do not directly affect any protective boundaries nor

do they impact the safety limits for the protective boundaries.

These proposed changes are administrative and editorial in nature.

Therefore, there can be 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: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

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

Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-

3499.

NRC Project Director: John F. Stolz

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

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

Connecticut

Date of amendment request: December 17, 1993

Description of amendment request: The proposed amendment would

revise the Technical Specifications to allow a relaxation in setpoint

tolerance of the pressurizer safety valves (PSVs) and main steam safety

valves (MSSVs) from plus or minus 1% to plus or minus 3% for the ``as-

found'' test condition.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

The proposed changes do not involve an SHC because the changes

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes revise the ``as found'' setpoint tolerances

for the PSVs and MSSVs from [plus or minus] 1% to [plus or minus]

3%. For the resetting of the PSVs and MSSVs, a [plus or minus] 1%

setpoint tolerance will be required prior to declaring the valve

operable for those instances where the [plus or minus] 1% tolerance

was exceeded. The proposed changes involve no hardware modifications

to plant structures, systems, or components. The proposed setpoint

tolerance of [plus or minus] 3% for the ``as-found'' condition was

previously evaluated as part of the PSE [Plant Safety Evaluation]

report for the transition to VANTAGE 5H fuel. The PSE was reviewed

and approved by the NRC staff as a part of a prior license

amendment.(9) In addition, since the proposed changes have

previously been evaluated by the PSE report, the calculated

radiological release associated with the PSE remain unaffected. In

addition, the proposed changes are in compliance with applicable

sections of the ASME Code and will not significantly affect

structural integrity of either the reactor coolant system or the

main steam system. Therefore, the proposed changes will have no

effect on the probability or consequences of previously evaluated

accidents.

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

from any accident previously evaluated.

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

different kind of accident from those previously analyzed. The

changes revise the Technical Specifications so that setpoint

tolerance for the PSVs and MSSVs can be [plus or minus] 3% for the

``as-found'' condition. These changes have no effect on plant

operation. The PSV and MSSV setpoint drift in excess of the [plus or

minus] 1% lift setting is an occurrence which has previously and may

subsequently occur. The analyses for the transition to the VANTAGE

5H fuel have examined the effects on the plant accident analyses for

relaxation in PSV and MSSV setpoint tolerance to [plus or minus] 3%.

Also, these changes will have no effect on ASME Code compliance.

These changes do not introduce any new failures.

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

In support of the transition to the VANTAGE 5H fuel, a PSE was

performed which assumed a [plus or minus] 3% setpoint tolerance for

both the PSVs and MSSVs. Therefore, the effects of relaxing the PSV

and MSSV setpoints are already accounted for in the existing

analyses of record and will not affect the plants accident analyses.

Additionally, the proposed changes will have no significant effect

on the structural integrity of the reactor coolant system or the

main steam system. Also, for those occurrences where the ``as-

found'' setpoint of the PSV or MSSV is in excess of [plus or minus]

1%, a resetting to within [plus or minus] 1% of the valve setpoint

will be required prior to declaring the valve operable. Therefore,

the proposed changes will not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

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

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

NRC Project Director: John F. Stolz

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

DiabloCanyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County, California

Date of amendment requests: December 14, 1993 (Reference LAR 93-07)

T3Description of amendment requests: The proposed amendment would

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

Power Plant (DCPP) Unit Nos. 1 and 2 to revise Technical Specification

(TS) 3/4.8.1, ``A.C. Sources'' to increase the required quantity of

emergency diesel generator (EDG) fuel oil stored in the engine-mounted

tank (day tank). The amendment request also proposes to revise TS 3/

4.7.11, ``Area Temperature Monitoring,'' and 3/4.8.1 to remove

references to a five EDG configuration. The specific TS changes

proposed are as follows:

(1) TS 3/4.7.11 would be revised to remove references to a common

(swing) diesel generator in Table 3.7-5.

(2) TS 3.8.1.1 and TS 3.8.1.2 would be revised to increase the

required minimum contained volume in the EDG engine-mounted fuel tank

(day tank) from 200 gallons to 250 gallons.

(3) TS 3.8.1.1 and TS 4.8.1.1.2 would be revised to remove

references to a five EDG configuration.

(4) TS 3.8.1.2 would be revised to correct a footnote. TS Bases 3/

4.8.1, 3/4.8.2, and 3/4.8.3 would be revised to clarify commitments to

Regulatory Guide 1.137 and expand the scope of information contained

within the TS Bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed increase in day tank TS minimum contained volume

requirements provides additional conservatism to assure the EDG fuel

oil contained in the day tank is sufficient to provide adequate time

for an operator to take corrective action to restore the fuel oil

supply to the affected day tank in the unlikely event that the fuel

oil supply from the main tanks were cut off.

Deletion of TS references to a five diesel generator

configuration and correction of the TS 3.8.1.2 footnote are

administrative changes that do not change the operating methodology

of DCPP. These proposed administrative changes remove outdated

information and correct an administrative oversight.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed increase in day tank TS minimum contained volume

requirements would not involve any physical change to the plant

systems or, in particular, to the EDG day tanks. The change does not

affect the ability of the EDGs to start and to fulfill their safety-

related function. Hence, no new failure mechanisms will be

introduced.

The proposed removal of references to a five EDG configuration

and correction of the TS 3.8.1.2 footnote are administrative in

nature. Further, the proposed changes would not result in any

physical alteration to any plant system. Therefore, the proposed

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

accident from any accident previously evaluated.

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

of safety?

Increasing the day tank TS minimum contained volume requirements

is a conservative change which provides additional margin to assure

the EDG fuel oil contained in the day tank is sufficient to provide

adequate time for an operator to take corrective action to restore

the fuel oil supply to the affected day tank in the unlikely event

that the fuel oil supply from the main tanks were cut off. The

proposed change will not alter any accident analysis assumptions,

initial conditions, or results. Consequently, the proposed change to

increase the EDG day tank TS contained fuel oil requirement does not

have any effect on the margin of safety.

The proposed administrative changes clarify the TS by removing

references to a five diesel generator configuration and correcting

the TS 3.8.1.2 footnote.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration. 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 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 and

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: December 9, 1993

Description of amendment request: The amendment would change the

Operating Licenses and their corresponding Appendices A to reflect the

planned implementation of the Power Rerate Program at Limerick

Generating Station Units 1 and 2, and the corresponding increase in the

authorized maximum reactor core power level by five percent to 3458

megawatts thermal (MWt) from the current limit of 3293 MWt.

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 Operation License (OL) changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed power rerate imposes only minor increases in the

plant operating conditions. Plant systems, components, and

structures have been verified to be capable of performing their

intended functions under rerated conditions. Where necessary, some

components will be modified or replaced prior to implementation of

the Power Rerate Program to accommodate the revised operating

condition. No new component or system interactions that could lead

to an accident are created. As discussed below, no transient events

result in a new sequence of events which could lead to a new

accident scenario. Emergency Core Cooling Systems (ECCS) - Loss-of-

Coolant Accident (LOCA) Analysis.

The current ECCS-LOCA performance analysis is already bounding

for power rerate conditions. The fuel peak cladding temperature for

rerate conditions is 1,345 deg.F, which is below the 2,200 deg.F

regulatory limit. Therefore, the analysis demonstrates that the LGS,

Units 1 and 2 will continue to comply with 10CFR50.46 and 10CFR50,

Appendix K.

Transient Event Analysis

The evaluation results for transient events indicate the margin

to the fuel Safety Limit Minimum Critical Power ratio (MCPR) will be

maintained for the 8x8 array fuel types, such as GE8x8NB or GE11

fuel design. The current fuel thermal-mechanical limits will

continue to be met.

Also, the power-dependent and flow-dependent MCPR and Maximum

Average Planar Linear Heat Generation Rate (MAPLHGR) limits

developed as part of the Average Power Range Monitor Rod Block

Monitor Technical Specifications (ARTS) improvement program are

applicable to power rerate. A TS Change Request to implement the

ARTS improvement program was submitted to the NRC by letter dated

August 27, 1993. The peak reactor vessel bottom head pressure will

remain within the American Society for Mechanical Engineers (ASME)

Code requirement for reactor overpressure protection.

The analysis performed focused on the most limiting transient

events in each disturbance category selected specifically for the

power rerate evaluations. The results demonstrated that LGS, Unit 1

and Unit 2 core thermal power output can be safely increased to

power rerate parameters without impacting plant safety during a

postulated transient event. The details of the impact to the

description in the UFSAR are delineated below.

a) Events Resulting in a Core Coolant Temperature Decrease

i) Loss of Feedwater Heating (LFWH)

The delta Critical Power Ratio (delta CPR) for the LFWH event at

the rerated power is bounded by the result estimated for the current

rated power level and remains significantly less than the Operating

Limit MCPR. There is no change between the delta CPR results for

high and low reactor core flow conditions. The calculated thermal

and mechanical overpowers for this event at power rerate conditions

also meet the fuel design criteria.

ii) Feedwater Controller Failure (FWCF) Maximum Demand

For the Increased Core Flow (ICF) and the Maximum Extended Load

Line Limit (MELLL) conditions, the trend for the FWCF - Maximum

Demand event at rerate conditions is consistent with the current

rated power analysis. For both high and low reactor core flow

conditions, the FWCF - Maximum Demand event becomes most limiting

due to the Turbine Bypass Valve Out-of-Service (TBVOOS) and the

Recirculation Pump Trip Out-of-Service (RPTOOS) analyses assumption.

The fuel thermal margin results remain within the acceptable limits

for the fuel type analyzed.

b) Events Resulting in a Reactor Pressure Increase

i) Turbine Trip with No Bypass (TTNBP)

At rerate conditions, the fuel transient thermal and mechanical

overpower results remain below the NRC acceptance criteria.

ii) Generator Load Rejection with No Bypass (LRNBP)

The fuel transient thermal responses are less severe than for

the TTNBP event described above. Therefore, at power rerate

conditions, the LRNBP event remains bounded by the TTNBP event.

iii) Main Steam Isolation Valve Closure, Flux Scram (MSIVF)

The peak reactor vessel bottom head pressure for rerate

conditions is slightly higher than the pressure at current rated

conditions due to the higher initial reactor coolant system

pressure. However, this result is still below the ASME overpressure

limit of 1,375 psig by a margin of 33 psi.

c) Events Resulting in a Core Coolant System Flow Rate Decrease

i) Recirculation Pump Seizure

The recirculation pump seizure assumes instantaneous stoppage of

the pump motor shaft of one recirculation pump. As a result, the

reactor core flow decreases rapidly. The reactor flow decreases

rapidly. The reactor vessel level swell due to the rapid reactor

core flow reduction reaches the high reactor water level setpoint,

causing a feedwater pump trip, a main turbine trip, and subsequently

a reactor scram on turbine stop valves closure. The peak neutron

flux and average fuel surface heat flux do not increase

significantly above the initial conditions, therefore no impact on

the fuel thermal margin is postulated to occur.

d) Events Resulting in Reactivity and Power Distribution

Anomalies

i) Rod Withdrawal Error (RWE)

The calculated delta CPR of 0.10 for this event at rerate

conditions is bounded by the generic ARTS - based RWE limits of

0.13. Therefore, the generic ARTS-based RWE analysis delta CPR

result is verified to be applicable for power rerate conditions for

LGS Units 1 and 2.

e) Events Resulting in a Reactor Coolant Inventory Increase

i) Inadvertent High Pressure Coolant Injection (HPCI) System

Actuation

Based on the peak average fuel surface heat flux results, the

HPCI actuation event will be bounded by the limiting pressurization

event (i.e., the TTNBP event described above) for delta CPR

consideration.

Anticipated Transients Without SCRAM (ATWS) Analysis

A generic evaluation for the ATWS event is provided in Section

3.7 of the Topical Report NEDC-31984P, ``Generic Evaluations of

General Electric Boiling Water Reactor Power Uprate,'' Supplement 1,

dated July 1991. This evaluation concludes that the ATWS acceptance

criteria for fuel, reactor pressure vessel (RPV) and containment

integrity will be met, if the following exists;

- Reactor power increases less than or equal to 5%

- Reactor Steam Dome pressure increases less than or equal to 40

psi;

- Safety Relief Valve (SRV) opening setpoints increase less than

or equal to 80 psi; and

- ATWS high pressure setpoint increases less than or equal to 20

psi.

The plant's parameter changes will remain within the above

criteria, except that the ATWS high pressure setpoint increase is 40

psi rather than 20 psi in order to maintain the same relationship

between the ATWS high pressure setpoint and the SRV opening

setpoints. Based on the previous analysis, this difference would

have a minor effect on the analysis results. The only significant

change is a slightly higher (i.e., about 10 psi) peak RPV pressure.

For additional assurance, a LGS specific ATWS analysis for a 5%

power rerate was performed. The events analyzed were:

1. Main Steam Isolation Valve (MSIV) Closure,

2. Pressure Regulator Failure - Open,

3. Loss of Feedwater, and

4. Inadvertent Opening of a Relief Valve.

The LGS specific analysis also concludes that the ATWS

acceptance criteria for fuel, RPV, and containment integrity will be

met for a 5% power rerate.

Other Evaluations

The impact of power rerate on the radiological consequences of

the accidents presented in UFSAR Chapter 15 was determined based on

the current design basis analyses, post rerate implementation system

conditions, and radiological source terms. In general, power rerate

will result in a small increase in the quantity of radioactive

material released during accidents and therefore slightly higher

(i.e., approximately 2% to 5%) accident doses. However, USFAR

Chapter 15 accident doses for rerated conditions remain within the

regulatory limits specified in 10CFR100 and 10CFR50, Appendix A, GDC

19.

The UFSAR Chapter 15 accidents that were evaluated and updated

for rerate conditions are as follows:

1) Loss of Coolant Accident (LOCA)

2) Main Steam Line Break (MSLB)

3) Fuel Handling Accident

4) Control Rod Drop Accident

5) Instrument Line Break

6) Feedwater Line Break

7) Steam Jet Air Ejector Line Break

8) Offgas System Failure

9) Liquid Radioactive Waste System Failure

An evaluation was also performed to address the power rerate

impact on accident mitigative features, structures, systems, and

components, within the balance of plant. The results are as follows:

- Auxiliary systems such as the Emergency Service Water,

Residual Heat Removal (RHR) Service Water, Ultimate Heat Sink (i.e.,

the spray pond), safety-related portions of secondary containment

reactor enclosure air cooling, primary containment drywell air

recirculation, and Emergency Diesel Generator enclosure ventilation

were confirmed to operate acceptably under normal and accident

conditions after implementation of power rerate.

- Combustible gas control systems were confirmed to be capable

of maintaining oxygen concentrations inside the primary containment

within regulatory limits under post accident rerate conditions.

- The secondary containment reactor enclosure recirculation

system and Standby Gas Treatment system were confirmed to be able to

adequately contain, process, and control the release of normal and

post-accident levels of radioactive material after implementation of

power rerate.

- Instrumentation was reviewed and confirmed to be capable of

performing their control and monitoring functions under rerate

conditions.

- Electric power systems including the main turbine generator

and switchgear components were verified as being capable of

providing the electrical load as a result of the rerated power

levels. No safety-related electrical loads were affected which would

impact the Emergency Diesel Generators.

- Piping systems were evaluated for the effect of operation at

higher power levels, including transient loadings. The evaluation

confirmed that with few exceptions piping and supports are adequate

to accommodate the increased loadings resulting from operation at

rerated power conditions. In a few cases, piping supports will be

modified to accept the higher forces due to rerate conditions.

- The effect of rerate conditions on high energy line break

(HELB) events for all Nuclear Steam Supply System (NSSS) and Balance

of Plant (BOP) systems was evaluated. The evaluation confirmed

structures, systems, and components important to safety are capable

of accommodating the effects of jet impingement and blowdown forces

and the environmental effects resulting from HELB events at rerate

conditions.

- The Moderate Energy Line Break (MELB) analysis was evaluated

for impact due to rerate conditions. Sufficient margin was

determined to exist in the original analysis to bound the rerate

conditions.

- Main control room (MCR) habitability was evaluated. Post-

accident MCR and Technical Support Center (TSC) doses were confirmed

to be within the limits of General Design Criterion (GDC) 19 of

10CFR50 Appendix A.

- Radiation doses for normal operation were reviewed and

confirmed to remain within the limits of 10CFR20 and 10CFR50,

Appendix I. The impact on post-accident sampling activities and

post-accident access to vital areas was also confirmed to be

acceptable.

- The environmental qualification of electrical and mechanical

equipment important to safety was evaluated for the impact of normal

and accident operating conditions at rerated power levels. The

majority of equipment will remain qualified for the new conditions.

For equipment that is not qualified, corrective actions will be

taken to ensure the plant equipment will perform their intended

functions under rerate conditions. No new equipment will be added

for power rerate which would increase the potential for component

failure. The Preventative Maintenance Program (PMP) will continue to

provide for appropriate equipment repair or replacement during

operation at rerated power conditions.

- The impact of operation at rerated power levels was evaluated

for Station Blackout and Fire Safety Shutdown area heat-up concerns.

The evaluation confirmed there is no adverse impact from rerate on

the ability of the plant to achieve safe shutdown under these

conditions.

- The consequences of postulated transients and special events

(i.e., ATWS and Station Blackout) will remain within NRC acceptance

criteria for rerate conditions. Concurrent malfunctions assumed to

occur during accidents have been accounted for in the safety

analyses for rerate conditions. The consequences of these equipment

malfunctions will not change with implementation of the Power Rerate

Program. Equipment that is important to safety either is capable of

or will be modified and/or replaced to be capable of performing its

intended function. The availability of redundant systems to provide

safety functions in the event of component malfunction is not

impacted as a result of rerate conditions. Furthermore, the impact

of power rerate on the consequences of abnormal transients and

accident conditions which are a result of component malfunctions has

been shown to be acceptable.

The probability (i.e., frequency of occurrence) of Design Basis

Accidents (DBAs) occurring is not affected by the proposed increased

power level, as the applicable regulatory criteria established for

plant equipment (e.g., ASME Code, the Institute of Electrical and

Electronics Engineers (IEEE) standards, National Electrical

Manufacturer's Association (NEMA) standards, NRC Regulatory Guides)

will still be followed as the plant is operated at the rerated power

level. Reactor SCRAM setpoints will be established such that there

is no significant increase in frequency due to rerate conditions. No

new challenges to safety-related equipment will result from the

implementation of power rerate.

The changes in consequences of hypothetical accidents which

would occur from 102% of the rerated power, compared to those

previously evaluated, are in all cases not significant, because the

accident evaluations from a power rerate to 105% of original rated

power will not result in exceeding the applicable NRC approved

acceptance limits. The spectrum of hypothetical accidents and

transients has been investigated, and has been determined to meet

the current regulatory criteria for LGS, Units 1 and 2 at rerate

conditions. The offsite radiological doses resulting from DBAs are

calculated to increase by only a few percent (i.e., approximately 2%

to 5%) because of the rerated power level, and will remain below

10CFR100 limits. In the area of reactor core design, the fuel

operating limits will continue to be met at the rerated power level,

and fuel reload analyses will continue to show that plant transients

will meet the criteria accepted by the NRC as specified in NEDO-

24011, ``GESTAR II.''

Challenges to fuel or ECCS performance were evaluated and shown

to still meet the criteria of 10CFR50.46 and 10CFR50, Appendix K.

Challenges to the primary containment have been evaluated and still

meet 10CFR50, Appendix A, GDC 38, ``Long Term Cooling,'' and GDC 50,

``Containment.'' Radiological release events have been evaluated and

have been shown to meet the guidelines of 10CFR100.

Therefore, the proposed OL changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2) The proposed OL changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

All actions to ensure that safety-related structures, systems,

and components will remain within their design allowable values, and

ensure that they can perform their intended functions under rerate

conditions will be taken prior to implementation of power rerate.

Power rerate does not increase challenges to or create any new

challenges to safety-related equipment or other equipment whose

failure could cause an accident. No new equipment is added as a

result of implementing the Power Rerate Program which would create

the possibility of a new type of accident. In addition, power rerate

does not create any new sequence of events or failure modes that

lead to a new type of accident.

Implementation of power rerate will increase the average neutron

flux in the reactor core, which increases the integrated neutron

fluence on the reactor pressure vessel (RPV) wall. To account for

the higher fluence, an RPV fracture toughness analysis was performed

for power rerate conditions. This analysis resulted in a proposed

revision to the ``pressure vs. temperature'' curves currently

provided in the Technical Specifications (TS), that will maintain

the current level of protection for the RPV. Therefore, power rerate

will not result in any new failure mode for the RPV, and thus, does

not create the possibility of a different type of accident from any

accident previously evaluated.

No new operating mode, safety-related equipment lineup, accident

scenario, or equipment failure mode was identified as resulting from

the implementation of the Power Rerate Program. The full spectrum of

accident considerations defined in NRC Regulatory Guide 1.70,

``Standard Format and Content of Safety Analysis Reports for Nuclear

Power Plants - LWR Edition,'' Revision 3, dated November 1978, have

been evaluated for rerate conditions and no new or different kind of

accident has been identified. Implementation of the Power Rerate

Program uses already-developed technology and applies it within the

capabilities of already existing plant equipment in accordance with

presently existing regulatory criteria to include applicable NRC

approved codes, standards, and methods. General Electric (GE) has

designed Boiling Water Reactors (BWRs) of higher power levels than

the rerated power of any of the currently operating BWR fleet and no

new power dependent accidents have been identified.

Therefore, the proposed OL changes do not create the possibility

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

evaluated.

3) The proposed OL changes do not involve a significant

reduction in a margin of safety.

Power rerate will not involve a significant reduction in a

margin of safety, as plant equipment and reactions to transients and

hypothetical accidents will not result in exceeding the presently

approved NRC acceptance limits. The accident doses are calculated to

increase a few percent (approximately 2% to 5%) because of power

rerate, but remain below 10 CFR 100 limits. The events (i.e.,

transients, accidents, and ATWS) that form the bases of the TS were

evaluated for power rerate conditions. Although some changes to the

TS are required to implement power rerate, no NRC acceptance limit

will be exceeded. Therefore, the margins of safety with respect to

the safety limits and other TS bases will be maintained.

For systems addressed in the TS Section 2.2, 3/4.1, 3/4.2, 3/

4.3, 3/4.4, 3/4.5, 3/4.6 and 3/4.7 (i.e., Reactor Protection System,

Standby Liquid Control System, Power Distribution Limits,

Instrumentation, Reactor Coolant System, Emergency Core Cooling

Systems, Containment Systems, and Plant Systems), all components

will be operable and capable of performing their intended functions

under power rerate conditions such that the margin of safety is not

adversely impacted.

Therefore, the proposed OL 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: 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: Charles L. Miller

Philadelphia Electric Company, Public Service Electric and Gas

Company,Delmarva Power and Light Company, and Atlantic City

Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom

Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: December 21, 1993

Description of amendment request: The proposed Technical

Specification (TS) changes revise Table 3.2.F, ``Surveillance

Instrumentation,'' to accurately describe the main stack high range and

reactor building roof vent high range radiation monitors, and deletes

previously approved TS Change Request (TSCR) 91-10 for Unit 3 (License

Amendment No. 168). TSCR 91-10 requested an emergency temporary change

to the TS to allow fuel loading to take place without all control rods

fully inserted into the 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:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Because the proposed changes are administrative in nature, they

do not affect the initial conditions or precursors assumed in the

Updated Final Safety Analysis Report Section 14. These changes do

not decrease the effectiveness of equipment relied upon to mitigate

the previously evaluated accidents.

Therefore, there is no increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any previously evaluated.

The proposed changes do not make any physical changes to the

plant or changes to operating procedures. Therefore, implementation

of the proposed changes will not affect the design function or

configuration of any component or introduce any new operating

scenarios or failure modes or accident initiation.

Therefore, the proposed changes do not create the possibility of

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

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes are administrative in nature and are

intended to provide clarification or eliminate confusion when

interpreting the Technical Specifications. The proposed changes do

not adversely affect the assumptions or sequence of events used in

any accident analysis.

Therefore, the proposed changes do not involve a reduction in

any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

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

Pennsylvania 17105

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: October 29, 1993

Description of amendment request: The licensee is preparing to

replace the currently installed steam generators with new model Delta

75 steam generators (Delta 75 SGs). The new steam generators will be

larger than those currently installed. The physical changes to the

plant and the accident reanalyses needed to support those changes will

necessitate changes to the Technical Specifications (TS). The TS

changes requested involve alterations to the core operating limits,

changes to various reactor trip setpoints, deletion of the negative

flux rate trip, removal of references to specific analyses, changes to

the steam/feedwater flow mismatch activation setpoint, changes to

shutdown limits, changes to instrument uncertainty allowances, a change

to the methodology for reactor coolant system (RCS) flow determination,

modifications to departure from nucleate boiling (DNB) parameters, a

change to the engineered safety features actuation system setpoints for

steam generator water levels, removal of the F* and L* criteria, and

the addition of a requirement for a first inservice inspection for the

new steam generators. Due to the size of the new steam generators, TS

containing references to the maximum containment pressure following a

steam line break and the total RCS volume will also change; in

addition, a reference to RCS temperature is changed from a nominal

value to an indicated value.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), South Carolina Electric

& Gas Company (SCE&G or the licensee) has provided its analysis of the

issue of no significant hazards consideration, which is presented

below:

1) Operation of VCSNS [Virgil C. Summer Nuclear Station] in

accordance with the proposed license amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Implementation of the [Delta] 75 SGs and revised operating

conditions do not contribute to the initiation of any accident

evaluated in the FSAR [Final Safety Analysis Report]. Supporting

factors are as follows:

- The [Delta] 75 SG is designed in accordance with ASME

[American Society of Mechanical Engineers] Code Section III, 1986

edition [sic] and other applicable federal, state, and local laws,

codes and regulations and meets the original interfaces for the

Model D3 SGs with exception that provisions for a larger blowdown

nozzle have been made and the feedwater inlet nozzle is located in

the upper shell.

- All NSSS [nuclear steam supply system] components (i.e.,

reactor vessel, RC Pumps, pressurizer, CRDM's [control rod drive

mechanisms], [Delta] 75 SGs, and RCS piping) are compatible with the

revised operating conditions. Their structural integrity is

maintained during all proposed plant conditions through compliance

with the ASME code.

- Fluid and auxiliary systems which are important to safety are

not adversely impacted and will continue to perform their design

function.

- Overall plant performance and operation are not significantly

altered by the proposed changes.

Therefore, since the reactor coolant pressure boundary integrity

and system functions are not adversely impacted, the probability of

occurrence of an accident evaluated in the VCSNS FSAR will be no

greater than the original design basis of the plant.

An extensive analysis has been performed to evaluate the

consequences of the following accident types currently evaluated in

the VCSNS FSAR:

- Non-LOCA [loss-of-coolant accident]

- Large Break LOCA

- Steam Generator Tube Rupture

With the [Delta] 75 SGs and revised operating conditions, the

calculated results (i.e., DNBR [departure from nucleate boiling

ratio], Primary and Secondary System Pressure, Peak Clad

Temperature, Metal Water Reaction, Challenge to Long Term Cooling,

Environmental Conditions Inside and Outside Containment, etc.) for

the accidents are similar to those currently reported in the VCSNS

FSAR. Select results (i.e., Containment Pressure During a Steam Line

Break, Minimum DNBR for Rod Withdrawal from Subcritical, etc.) are

slightly more limiting than those reported in the current FSAR due

to the use of the assumed operating conditions with the new [Delta]

75 SGs, and in some cases, use of an uprated core power of 2900 MWt.

However, in all cases, the calculated results do not challenge the

integrity of the primary/secondary/ containment pressure boundary

and remain within the regulatory acceptance criteria applied to

VCSNS's current licensing basis. The assumptions utilized in the

radiological evaluations, described in Section 3.7, are thus

appropriate and are judged to provide a conservative estimate of the

radiological consequences during accident conditions. Given that

calculated radiological consequences are not significantly higher

than current FSAR results and remain well within 10CFR100 limits, it

is concluded that the consequences of an accident previously

evaluated in the FSAR are not increased.

2) The proposed license amendment does not create the

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

previously evaluated.

The [Delta] 75 SGs and revised operating conditions will not

introduce any new accident initiator mechanisms. Structural

integrity of the RCS is maintained during all plant conditions

through compliance with the ASME code. No new failure modes or

limiting single failures have been identified. Design requirements

of auxiliary systems are met with the RSGs [Replacement Steam

Generators]. Since the safety and design requirements continue to be

met and the integrity of the reactor coolant system pressure

boundary is not challenged, no new accident scenarios have been

created. Therefore, the types of accidents defined in the FSAR

continue to represent the credible spectrum of events to be analyzed

which determine safe plant operation.

3) The proposed license amendment does not involve a significant

reduction in a margin of safety.

Although the [Delta] 75 SGs and revised operating conditions

will require changes to the VCSNS Technical Specifications, it will

not invalidate the LOCA, non-LOCA, or SGTR [steam generator tube

rupture] conclusions presented in the FSAR accident analyses

(Appendix 6). For all the FSAR non-LOCA transients, the DNB design

basis, primary and secondary pressure limits, and dose limits

continue to be met. The LOCA peak cladding temperatures remain below

the limits specified in 10CFR50.46. The calculated doses resulting

from a SGTR event will continue to remain within a small fraction of

the 10CFR100 permissible releases. Environmental conditions

associated with High Energy Line Break (HELB) both inside and

outside containment have been evaluated. The containment design

pressure will not be violated as a result of the HELB. Equipment

qualification will be updated, as necessary, to reflect the revised

conditions resulting from HELB. The margin of safety with respect to

primary pressure boundary is provided, in part, by the safety

factors included in the ASME Code. Since the components remain in

compliance with the codes and standards in effect when VCSNS was

originally licensed (with the exception of the [Delta] 75 RSGs which

use the 1986 ASME Code Section III Edition), the margin of safety is

not reduced. Thus, there is no reduction in the margin to safety as

defined in the bases of the VCSNS Technical Specifications.

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: S. Singh Bajwa

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: December 17, 1993

Description of amendment request: The proposed changes would revise

Technical Specification 3/4.3.3.6, ``Accident Monitoring

Instrumentation,'' and the associated Technical Specification Bases.

The changes are in accordance with the applicable guidance of Revision

3 to Regulatory Guide (RG) 1.97.

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 would

not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Regulatory Guide 1.97 furnishes standards acceptable to the NRC

for instrumentation to monitor plant variables and systems during

and following an accident. The purpose of the accident monitoring

instrumentation is to display plant variables that provide

information required by the control room operators for manual

actions and long term recovery. Determination of variable types and

category designations for VCSNS [Virgil C. Summer Nuclear Station]

was accomplished from a review of the Emergency Response Guidelines

(ERGs), the Final Safety Analysis Report, and the Westinghouse

Owners Group (WOG) ERGs. The WOG ERGs were used at VCSNS as a basis

for the Emergency Response Procedures. Operability of the

instruments used for accident monitoring ensures there is sufficient

information available on selected plant parameters to monitor plant

status during and following an accident. The changes proposed do not

effect components that can cause an accident. The increase in

allowable outage times from 7 to 30 days or from 48 hours to 7 days

does not significantly affect the consequences of an event

previously evaluated. The channel redundancy and the relatively

short outage times, coupled with the low probability of an event

requiring accident monitoring instrumentation during this interval,

ensure that sufficient information is available for operator manual

actions. The condition of the plant in either HOT STANDBY or HOT

SHUTDOWN, the first stage of the plant shutdown process, has no

impact on the assumptions made in the accident analysis.

The change in mode applicability for the Reactor Building Area

High Range Radiation Monitors to include modes 1, 2, and 3, but

exclude mode 4, is based on the usage of these monitors which is to

indicate a significant degradation of the reactor coolant pressure

boundary. These monitors do not initiate any automatic mitigation

system and are solely required to be operable to provide indication

which in conjunction with other operator actions will aid in

mitigating the consequences of design basis accidents. Design basis

accident sequences which may create a significant degradation of the

reactor coolant pressure boundary are not postulated to occur during

mode 4. Therefore, the proposed change does not increase the

probability or consequences of any accident previously evaluated.

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

accident from any previously evaluated.

The proposed change is consistent with the requirements of RG

1.97. The accident monitoring instrumentation will make available

reliable information to plant control room operators to mitigate the

consequences of a design basis accident. The first stages of plant

shutdown, HOT STANDBY and HOT SHUTDOWN, are plant modes for which

VCSNS has been analyzed. Since no plant configuration changes or

changes to the mode of operation of equipment, systems, and

components are introduced by the proposed Technical Specification,

no new failure modes or accident sequences are instituted.

Therefore, the changes proposed do not create the possibility of a

new or different kind of accident from any previously analyzed.

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

The inclusion of category 1, type A or B, instrumentation in the

TS [Technical Specifications] provides assurance that adequate

information is available to the operators to maintain VCSNS in a

safe condition during and following a design basis accident.

Accomplishment of specific manual action by the control room

operators is enhanced due to the availability and reliability of the

indications. The proposed changes do not affect the design or

operation of safety related components relied upon to automatically

mitigate the consequences of a design basis event. The proposed

change from HOT SHUTDOWN to HOT STANDBY as the first stage of plant

shutdown will not affect the design or operation of any safety

related system or component. Therefore, the changes proposed would

not involve a reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: S. Singh Bajwa

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: November 3, 1993

T3Description of amendment requests: The licensee proposes to

revise the operability requirements of containment isolation valves

listed in Technical Specification (TS) Table 3.6-1, Section D. The

associated Bases 3/5.6.3, ``Containment Isolation Valves,'' is also

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:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No

The proposed change provides new actions and Allowed Outage

Times (AOTs) for valves in Section D of Technical Specification (TS)

Table 3.6-1 that are currently allowed by the existing TS to be

secured for an indefinite period of time as long as they are secured

in their Engineered Safety Feature Actuation System (ESFAS) actuated

position. These valves are considered operable by the existing TS

although they may be unable to perform their containment isolation

function. The proposed change ensures that these valves are returned

to operable status within specified times based on the results of

specific risk evaluations on their contribution to core damage or

offsite dose release. The proposed change does not involve a

physical change to the facility as described in the Updated Final

Safety Analysis Report (UFSAR). Therefore, this proposed change does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

The ESFAS actuated positions of these valves are the positions

assumed in the safety analysis. There are no new accidents

associated with this proposed change because the previously analyzed

events already considered failures of containment isolation valves.

The plant is equipped with dual and redundant containment isolation

valves. Leaving the valves in their ESFAS actuated positions does

not create a new accident. Therefore, this proposed change 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

This proposed change 1) limits the AOT of certain valves based

on contributions to core damage and offsite dose release when the

valves are secured in their ESFAS actuated position and 2) requires

these valves to be returned to OPERABLE status prior to Mode 4 entry

from a cold shutdown to ensure they are available to perform their

intended containment isolation function. Previously, these valves

could be secured in the ESFAS actuated position indefinitely.

Therefore, this proposed change does not involve a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 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

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: December 27, 1993

Description of amendment request: The proposed change would revise

the Technical Specifications (TS) for the North Anna Power Station,

Units No. 1 and No. 2 (NA-1&2). The proposed changes revise the review

responsibilities of the Station Nuclear Safety and Operating Committee

(SNSOC) and the Management Safety Review Committee (MSRC).

The NA-1&2 TS address the organization and responsibilities of both

the onsite and offsite review groups: SNSOC and 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. The extent of these review activities

would be revised by the proposed changes to ensure the two review

groups are focusing on nuclear safety issues and not spending an

unnecessary amount of time on activities of minimal safety

significance. Specifically, the proposed changes would revise the

review responsibilities of SNSOC regarding procedure changes. Rather

than reviewing all procedure changes, SNSOC would only review procedure

changes that require a safety evaluation. The proposed changes also

would revise the review responsibilities of the MSRC. Rather than

reviewing all of the safety evaluations and SNSOC meeting minutes and

reports as presently required by the TS, the MSRC would only review a

representative sample of these documents.

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:

[T]he elimination of the SNSOC review of procedure changes that

do not require a safety evaluation, revising the wording for

approval of procedure changes, and the modification of the MSRC's

duties regarding their review of safety evaluations and SNSOC

meeting minutes and reports 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

administative procedures, however, only changes that require a

safety evaluation will require SNSOC approval. MSRC review of

representative 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 location: 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: Herbert N. Berkow

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.

Duke Power Company, Docket No. 50-413, Catawba Nuclear Station,

Unit No. 1, York County, South Carolina

Date of amendment request: January 10, 1994

Description of amendment request: The proposed amendment would

revise Technical Specifications 2.0 and 3/4.2 which currently requires

the determination of the reactor coolant system flow rate by precision

heat balance measurement at least once per 18 months. Date of

publication of individual notice in Federal Register: January 26, 1994

(59 FR 3743)

Expiration date of individual notice: February 25, 1994

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of amendment request: November 11, 1993, as supplemented

November 22, 1993

Description of amendment request: The proposed amendments would

provide an interim acceptance criteria for control rod drop time on

Oconee Unit 1. Specifically, control rod Group 1, Rod 8, and Group 2,

Rod 5, would be considered operable with an insertion time of less than

or equal to 3.00 seconds provided that: (1) the average insertion time

for the remaining rods in Group 1 and the average insertion time for

the remaining rods in Group 2 is less than or equal to 1.5 seconds, and

(2) the core average negative reactivity insertion rate is within the

assumptions of the safety analysis. The acceptance criteria would apply

until the end of the current fuel cycle for Oconee Unit 1. This

acceptance criteria for rod drop time would apply for the two rods,

rather than the existing Technical Specification 4.7.1 limit of 2.00

seconds from the fully withdrawn position to 3/4 insertion.Date of

publication of individual notice in Federal Register: November 29, 1993

(58 FR 62689)

Expiration date of individual notice: December 29, 1993

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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 application for amendment: December 8,

1993I11T3Brief description of amendment request: The

proposed amendment would grant one-time extensions for certain

Technical Specification surveillances which are currently required to

be performed beginning February 16, 1994. The licensee is requesting

extension of the surveillance intervals because the current operating

cycle has been extended, impacting the required completion dates for

these surveillances. Performance of these surveillances within the

required intervals would require that the plant be placed in an

undesirable operating configuration, or would necessitate a plant

shutdown. The surveillances for which extensions have been requested

will be performed during the fifth refueling outage, scheduled to begin

on April 16, 1994.

Date of individual notice in Federal Register: January 18, 1994 (59

FR 2630)

Expiration date of individual notice: February 17, 1994

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

Power Authority of the State of New York, Docket No. 50-333, James

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: December 22, 1993

Description of amendment request: The proposed amendment would add

Limiting Conditions for Operation (LCO) and Surveillance Requirements

to Tables 3.12.1, ``Water Spray/Sprinkler Protected Areas,'' and

4.12.1, ``Water Spray/Sprinkler Tests,'' and clarify the associated

Bases to reflect the installation of a new full area fire suppression

system in the east and west cable tunnels. This new full area fire

suppression system was installed because the previous sprinkler system

did not provide coverage to some cable trays and the sprinkler head

orientation did not provide full coverage of the cable trays where it

was installed. The proposed amendment would also correct other portions

of Tables 3.12.1 and 4.12.1 for consistency with changes made to

reflect the east and west cable tunnel modification.Date of publication

of individual notice in Federal Register: January 18, 1994 (59 FR 2634)

Expiration date of individual notice: February 17, 1994

Local Public Document Room location: Reference and Documents

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

York 13126.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.

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: October 28, 1992, as

supplemented December 14, 1993

Brief description of amendments: The amendments remove Table 4.4-5,

``Reactor Vessel Material Surveillance Program Withdrawal Schedule,''

from the McGuire Technical Specifications and make other administrative

changes associated with the removal of the withdrawal schedule in

accordance with NRC Generic Letter 91-01.

Date of issuance: January 31, 1994

Effective date: January 31, 1994

Amendment Nos.: 139 and 121

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 23, 1992 (57

FR 61112) The December 14, 1993, letter provided clarifying information

that did not change the scope of the October 28, 1992, 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 January 31, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Astkins Library, University of

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

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

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

CarolinaDate of application of amendments: July 14, 1993, as

supplemented August 24 and September 22, 1993

Brief description of amendments: The amendments revise TS 3.1.2.9

to clarify the role of High Pressure Injection and Core Flood Tank

deactivation in maintaining pilot operated relief valve operability for

low temperature overpressure protection (LTOP), add restrictions

regarding applicability of controls which assure 10 minutes are

available for operator action to mitigate an LTOP event, revise the

pressure-temperature limits and associated LTOP setpoints, and make

associated administrative changes. Also, the Bases would be revised to

be consistent with the above changes.

The conformance of the upper shelf energy and reactor vessel

material surveillance program to Appendices G and H will be determined

pending the NRC staff resolution of Generic Letter 92-01 in 1994.

Date of issuance: January 25, 1994

Effective date: To be issued within 30 days from the date of

issuance

Amendment Nos.: 204, 204, and 201

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

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 1, 1993 (58

FR 46228) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 25, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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: February 19, 1993

Brief description of amendments: The amendments revise the Appendix

A TSs 3.4.9.1, 3.4.9.2, and 4.4.9.2 relating to pressurizer surge line

stratification.

Date of issuance: January 31, 1994

Effective date: January 31, 1994

Amendment Nos.: 179 and 59

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

revised the Technical Specifications.

Date of initial notice in Federal Register: Aspril 28, 1993 (58 FR

25854) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 31, 1994.No 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. 50-368, Arkansas Nuclear

One,Unit No. 2, Pope County, Arkansas

Date of application for amendment: February 24, 1993

Brief description of amendment: The amendment revised the

containment internal pressure lower limit of Technical Specification

Figure 3.6-1 from 12.8 to 13.2 psia.

Date of issuance: February 3, 1994

Effective date: 30-days from date of issuance

Amendment No.: 156

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

Technical Specifications.

Date of initial notice in Federal Register: March 31, 1993 (58 FR

16858) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 3, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

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

Date of application for amendments: October 4, 1993

Brief description of amendments: The amendments revise the

surveillance test schedule in TS 4.6.1.2a and the associated Bases for

performing Type A test which determine the overall integrated

containment leakage rate.

Date of issuance: January 11, 1994

Effective date: January 11, 1994Amendment Nos. 158, 152Facility

Operating Licenses Nos. DPR-31 and DPR-41: Amendments revised the

Technical Specifications.

Date of initial notice in Federal Register: November 10, 1993 (58

FR 59748) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 11, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket No. 50-498, South Texas Project, Unit 1, Matagorda County,

Texas

Date of amendment request: December 6, 1993

Brief description of amendments: The amendments modify Technical

Specification 3.7.1.2 by extending the allowed outage time for the Unit

1 Train D turbine-driven auxiliary feedwater pump from 72 hours to 168

hours. This change is a one-time-only extension to accommodate an

augmented test program for the turbine driven auxiliary feedwater pump

during the restart of Unit 1 from the 1993 outage.

Date of issuance: January 25, 1994

Effective date: January 25, 1994, to be implemented within 10 days

of issuance.

Amendment No.: 58

Facility Operating License No. NPF-76. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 22, 1993 (58

FR 67848). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 25, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

Illinois Power Company and Soyland Power Cooperative, Inc., Docket

No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,

Illinois

Date of application for amendment: November 4, 1993

Brief description of amendment: The amendment revises Clinton

Technical Specification 3/4.8.1.1, ``AC Sources - Operating,'' by

relocating the surveillance requirement to inspect the diesel

generators in accordance with the manufacturer's recommendations to the

preventive maintenance program.

Date of issuance: January 31, 1994

Effective date: January 31, 1994

Amendment No.: 87

Facility Operating License No. NPF-62. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 8, 1993 (58 FR

64610) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 31, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: The Vespasian Warner Public

Library District, 310 N. Quincy Street, Clinton, Illinois 61727.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket

No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,

IllinoisDate of application for amendment: November 4, 1993

Brief description of amendment: The licensee proposed modifying

Technical Specification 3/4.8.2.1, ``DC Sources - Operating,'' by

deleting the requirement that the plant be shut down to perform the

required battery capacity or service testing. Following discussions

with the licensee, the staff has modified the licensee's proposal and

approved a one-time only change to permit replacement of the Division

IV battery subsystem at power.

Date of issuance: February 2, 1994

Effective date: February 2, 1994

Amendment No.: 88

Facility Operating License No. NPF-62. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 8, 1993 (58 FR

64610) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 2, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: The Vespasian Warner Public

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

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook,

Nuclear Plant, Unit No. 2, Berrien County, Michigan

Date of application for amendment: Aspril 16, 1993, as supplemented

September 28 and December 3, 1993

Brief description of amendment: The amendment revises Technical

Specifications to allow certain tests normally designated as 18-month

surveillances to be delayed until the next refueling outage scheduled

to begin August 6, 1994. Extensions for four groups of surveillances

(Groups 1, 2, 6, 11) were previously approved for Unit 2 in Amendment

158 dated December 22, 1993. This amendment grants approval for the

extensions requested for the remaining 12 groups of surveillances and

completes the staff's review of the licensee's April 16, 1993 (as

supplemented) application.

Date of issuance: January 26, 1994

Effective date: January 26, 1994

Amendment No.: 159

Facility Operating License No. DPR-74. Amendment revises the

Technical Specifications. Dates of initial notice in Federal Register:

Asugust 4, 1993 (58 FR 41505) and December 21, 1993 (58 FR 67850)The

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

Safety Evaluation dated January 26, 1994. No significant hazards

consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085.Northeast

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

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

Date of application for amendment: June 11, 1993, supplemented by

letter dated November 15, 1993

Brief description of amendment: The amendment revises the pressure/

temperature (P/T) limits for the reactor vessel. Specifically, Figure

3.4-2, ``Millstone Unit 2 Reactor Coolant System Presure-Temperature

Limitations for 12 Full Power Years,'' on page 3/4 4-19, is revised to

reflect the change in the curves and the title change to ``Millstone

Unit 2 Reactor Coolant System Pressure-Temperature Limitations for 20

EFPY.''

Date of issuance: January 27, 1994

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

within30 days.

Amendment No.: 170

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

Technical Specifications.

Date of initial notice in Federal Register: July 21, 1993 (58 FR

39054) The November 15, 1993, submittal provided information that did

not change the initial proposed no significant hazards consideration

determination. The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 27, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, Connecticut 06360.

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

DiabloCanyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County,California

Date of application for amendments: December 22, 1992, as

supplemented July 19, 1993 (Reference LAR 92-08)

Brief description of amendments: The amendments revise the combined

Technical Specifications (TS) for the Diablo Canyon Power Plant Unit

Nos. 1 and 2. Specifically, TS Section 3/4.3.2, ``Engineered Safety

Features Actuation System Instrumentation,'' would be revised to change

the second level undervoltage trip setpoint and allowable values.

Technical Specification 3/4.8.1, ``A.C. Sources,'' would also be

changed to revise the diesel generator (DG) steady state voltage

surveillance requirements. The second level undervoltage relay TS

setpoint and allowable values will be changed to maintain acceptable

voltages at the 480 volt and 120 volt buses during sustained degraded

voltage conditions. The DG steady state voltage surveillance

requirements will be changed to ensure that the diesel generators

provide adequate voltage when required to power the vital loads.

Date of issuance: January 6, 1994

Effective date: 60 days from date of issuance

Amendment Nos.: 86 & 85

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 3, 1993 (58 FR

7002) The July 19, 1993 submittal provided clarifying information and

did not affect the initial Federal Register notice and proposed no

significant hazards consideration. The Commission's related evaluation

of the amendments is contained in a Safety Evaluation dated January 6,

1994No significant hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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 application for amendments: July 6, 1993, as supplemented

December 29, 1993 (Reference LAR 93-03)

Brief description of amendments: The amendments revise the combined

Technical Specifications (TS) 3/4.3.2, ``Engineered Safety Features

Actuation System Instrumentation,'' Table 4.3-2, ``Engineered Safety

Features Actuation System Instrumentation Surveillance Requirements,''

for the Diablo Canyon Power Plant Unit Nos. 1 and 2 to relax the slave

relay test frequency for slave relays K612A, K614B, K615A, and K615B

from quarterly to once per 18 months during refueling or extended cold

shutdowns. The affected slave relays cause isolation of the charging

and letdown portions of the chemical and volume control system, and

actuate charging pump suction valves associated with volume control

tank and refueling water storage tank isolation.

Date of issuance: January 31, 1994

Effective date: For cycle 7 and after

Amendment Nos.: 87 and 86

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: Asugust 18, 1993 (58 FR

43929) The December 29, 1993, submittal provided clarifying information

and did not effect the initial Federal Register Notice and proposed no

significant hazards consideration.The Commission's related evaluation

of the amendments is contained in a Safety Evaluation dated January 31,

1994.No significan hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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

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

Pennsylvania

Date of application for amendments: January 9, 1991, as

supplemented on August 19, 1991, June 22, 1992 and August 3, 1992

Brief description of amendments: The amendment changed the

Technical Specifications to revise the isolation setpoints for the

ambient temperature switches for the High Pressure Coolant Injection

and Reactor Core Isolation Cooling Systems room area coolers.

Date of issuance: January 31, 1993

Effective date: January 31, 1993

Amendment Nos.: 132 and 99

Facility Operating License Nos. NPF-14 and NPF-22. These amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 9, 1993 (58 FR

32389) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 31, 1993.No significant

hazards consideration comments received: No

Local Public Document Room location: Osterhout Free Library,

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

Pennsylvania 18701.

Philadelphia Electric Company, Docket No. 50-353, Limerick

Generating Station, Unit 2, Montgomery County, Pennsylvania

Date of application for amendment: Asugust 27, 1993, as

supplemented November 10, and December 20, 1993

Brief description of amendment: The amendment allows a one-time TS

change to extend the allowed outage time (AOTs) for the Unit 2 residual

heat removal service water (RHRSW) system as well as the suppression

pool spray and suppression pool cooling modes of the residual heat

removal system from 72, 168 (i.e. seven days), and 72 hours,

respectively, to 288 hours (i.e., twelve days). The extended AOTs would

allow continued Unit 2 operation while maintenance isolation valves are

installed on both loops of the RHRSW system.

Date of issuance: January 26, 1994

Effective date: January 26, 1994

Amendment No. 30

Facility Operating License No. NPF-85. This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 29, 1993 (58

FR 50970) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 26, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Public Service Electric and Gas

CompanyDelmarva Power and Light Company, and Atlantic City Electric

Company,Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power

Station,Unit Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: October 5, 1993

Brief description of amendments: This amendment revised the Plant

Operating Review Committee review, the Nuclear Review Board review, the

Radiological Environmental Monitoring Program requirements, position

titles, and the organization chart in Appendix B of the Technical

Specifications (TS) to be consistent with Appendix A of the TS.

Date of issuance: January 26, 1994

Effective date: January 26, 1994Amendments Nos.: 183 and 188

Facility Operating License Nos. DPR-44 and DPR-56: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 8, 1993 (58 FR

64612) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 26, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

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

Pennsylvania 17105.

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

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: February 2, 1993, and

supplemented by letter dated November 16, 1993.

Brief description of amendment: The amendment extends the period of

time to reduce the setpoints of the Average Power Range Monitors and

the Rod Block Monitor when the plant enters single-loop operations.

Additionally, the change incorporates updated core values relative to

single loop operations and the addition of a new Specification 3.0.5

and its associated Bases.

Date of issuance: January 25, 1994

Effective date: As of date of issuance and shall be implemented

within 60 days of the date of issuance.

Amendment No.: 63

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

Technical Specifications.

Date of initial notice in Federal Register: March 31, 1993 (58 FR

16872) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 25, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: May 21, 1993, as supplemented on

October 29, 1993, and November 16, 1993; the staff's proposed finding

of no significant hazards is not affected by these supplements.

Brief description of amendment: This amendment revises Technical

Specifications surveillance requirement 4.4.2.2 to apply only to the

pilot stage assembly of the safety relief valves (SRVs) and adds a new

surveillance requirement which will require the main portion of the

SRVs to be set pressure tested at least once every 5 years.

Date of issuance: January 27, 1994

Effective date: January 27, 1994

Amendment No.: 64

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

Technical Specifications.

Date of initial notice in Federal Register: Asugus

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