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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice Applications and Amendments to Facility Operating

Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from December 13, 1993, through December 22,

1993. The last biweekly notice was published on December 22, 1993 (58

FR 67840).

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 February 4, 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.

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

50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit

Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: December 2, 1993

Description of amendment requests: The proposed changes would

modify TS 3/4.6.1.2 by removing the schedular requirements for a Type A

(overall integrated containment leakage rate) test to be performed

specifically at 40 1B 10 month intervals and replacing these

requirements with a requirement to perform Type A testing in accordance

with Appendix J to 10 CFR 50.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes would not involve an increase in the

probability or the consequences of an accident previously evaluated.

The proposed change only allows flexibility in the scheduling of the

three required Type A tests in the 10-year service period. The

additional flexibility is needed for plants using 18-month fuel

cycles to allow refueling outages and testing intervals to coincide.

There is no change to the number of tests required, test

methodology, or acceptance criteria.

(2) The proposed changes would not create the possibility of a

new or different type of accident from any accident previously

evaluated. The proposed change to the test schedule only provides

flexibility in meeting the same requirement for three tests in a 10-

year period. The testing type and bases have not changed. Therefore,

operation of the units with this more flexible test schedule will

not result in an accident previously not analyzed in the Updated

Final Safety Analysis Report (UFSAR). The proposed changes do not

impact the design bases of the containment and do not modify the

response of the containment during a design basis accident.

(3) The proposed changes would not involve a reduction in the

margin of safety. The proposed changes to the schedule only provides

flexibility in meeting the same requirement for three tests in a 10-

year period. These proposed changes do not affect or change any

limiting conditions for operation (LCO), or any other surveillance

requirements in the TS, and the basis for the surveillance

requirement remains unchanged. The testing method, acceptance

criteria, and bases are not changed. The TS continue to require

testing that is consistent with the requirements of Appendix J to 10

CFR 50.

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

that 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: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: Theodore R. Quay

Gulf States Utilities Company, Docket No. 50-458, River Bend

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: November 18, 1993

Description of amendment request: The proposed amendment would

permit extending the time to perform leak rate testing of certain

containment isolation valves so that the testing can be performed

during the refueling outage scheduled to start April 16, 1994, rather

than requiring an earlier shutdown solely to perform the testing. The

proposed amendment would revise Surveillance Requirements 4.6.1.3d and

4.6.1.3f to allow a one-time extension of the surveillance intervals

for leak rate testing of containment isolation valves. In addition, the

proposed amendment would revise Surveillance Requirements 4.4.3.2.2a

and 4.4.3.2.2b, replacing the requirement to leak test the reactor

coolant pressure isolation valves every 18 months or prior to returning

a valve to service, with a requirement to leak test the valves in

accordance with the Inservice Testing Program. This would allow the

testing to be performed during the fifth refueling outage.

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. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's review is

presented below:

1. The proposed changes would not significantly increase the

probability or consequences of a previously evaluated accident.

One of the proposed technical specification (TS) changes requests a

one-time only extension of the surveillance intervals for the TS

Surveillance Requirements of TS 4.6.1.3f, leak rate testing of valves

sealed by the main steam positive leakage control system (MS-PLCS) and

the penetration valve leakage control system (PVLCS). The revision

would permit eleven containment isolation valves to be tested a maximum

of 46 days later than required by current technical specifications.

To permit the one-time extension of the surveillance interval for

leak rate tests of containment isolation valves, TS 4.6.1.3d must also

be revised to permit the interval for Type C leak rate tests to exceed

24 months. This change is consistent with an associated exemption

request. The exemption request and this revision would permit 20 valves

to be tested a maximum of 35 days later than required by the current

technical specifications.

The proposed amendment would also revise Surveillance Requirements

4.4.3.2.2a and 4.4.3.2.2b, replacing the requirement to leak test the

reactor coolant pressure isolation valves every 18 months or prior to

returning a valve to service, with a requirement to leak test the

valves in accordance with the Inservice Testing Program. This change

would require that the pressure isolation valves be tested in

accordance with Section XI of the ASME Boiler and Pressure Vessel Code,

resulting in the valves being tested at least every refueling outage,

rather than specifying an 18 month cycle. The revision would permit

five valves to be tested a maximum of 65 days later than allowed under

the current technical specification.

Based on the short duration of the requested extensions, the

extensions will not significantly increase the probability or

consequences of a previously evaluated accident.

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

or different kind of accident from any previously evaluated.

The proposed TS changes permit extension of the surveillance

intervals for leak rate testing of containment isolation valves and

reactor coolant system pressure isolation valves. In that the requested

extension durations are small as compared to the overall interval

allowed by TS, the proposed changes do not create the possibility of a

new or different kind of accident from any accident previously

analyzed.

3. The proposed changes would not involve a significant reduction

in the margin of safety.

The proposed TS changes permit extension of the surveillance

intervals for leak rate testing of containment isolation valves and

reactor coolant system pressure isolation valves. In that the requested

extension durations are small as compared to the overall interval

allowed by TS, the proposed changes do not involve a significant

reduction in the margin of safety.

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 Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005

NRC Project Director: Suzanne C. Black

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendments request: December 8, 1993

Description of amendments request: The frequency for Channel

Calibration would be revised from Q (quarterly) to R (refuel) for

Technical Specification Table 4.3.2.1, Item 4.a.4, High Pressure Core

Injection Steam Line Tunnel Temperature-High.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The proposed change corrects Technical

Specification pages issued for Amendment 166 for Brunswick Unit 1

and Amendment 197 for Brunswick Unit 2, regarding NUMAC Steam Leak

Detection Equipment. Specifically, on page 3/4 3-29 for each unit,

the Channel Calibration frequency of Item 4.a.4, HPCI [High Pressure

Core Injection] Steam Line Tunnel Temperature - High, was

inadvertently left as quarterly (Q) rather than being revised to

refuel (R). The text of CP&L's September 14, 1992 license amendment

request and the NRC's safety evaluation for Amendments 166 and 197,

dated October 14, 1993, addressed the frequency change from

quarterly to refuel for this item. Therefore, the proposed change is

purely administrative in nature and can not involve a significant

increase in the probability of consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. As stated above, the NRC's safety evaluation for

Amendments 166 and 197, dated October 14, 1993, addressed the

frequency change from quarterly to refuel for Item 4.a.4 of Table

4.3.2-1, HPCI Steam Line Tunnel Temperature - High. Therefore, the

proposed change is purely administrative in nature and can not

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

any accident previously evaluated.

3. The proposed license amendment does not involve a significant

reduction in the margin of safety. The proposed change corrects

Technical Specification pages issued for Amendment 166 for Brunswick

Unit 1 and Amendment 197 for Brunswick Unit 2, regarding NUMAC Steam

Leak Detection Equipment. The NRC's safety evaluation for Amendments

166 and 197, dated October 14, 1993, addressed the change of Channel

Calibration frequency of Item 4.a.4, HPCI Steam Line Tunnel

Temperature - High from quarterly (Q) to refuel (R). Therefore, the

proposed change is purely administrative and can not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

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

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

NRC Project Director: S. Singh Bajwa

Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling

Water Reactor (LACBWR), Vernon County, Wisconsin

Date of application for amendment: November 5, 1993 (Reference LAC-

13320)

Brief description of amendment: This proposed change would modify

the Technical Specifications incorporated in Facility Operating License

No. DPR-45 in accordance with the requirements of the revised 10 CFR

Part 20 which becomes mandatory January 1, 1994 (56 FR 23360). In

addition, this proposed change would correct several editorial

oversights from previous amendments.

Basis for proposed no significant hazards consideration

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

the results of its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the information provided by

the licensee and found that the licensee did not provide specific

information as to how it determined that the three standards of

50.92(c) were satisfied. The NRC staff performed its own evaluation of

the proposed change to determine if the three standards of 50.92(c)

were satisfied. The NRC staff's no significant hazards consideration

evaluation is presented below:

1. Will operation of the facility according to this proposed change

involve a significant increase in the probability or consequences of an

accident previously evaluated?

The proposed change is to bring the LACBWR Technical Specifications

into conformance with the revised 10 CFR Part 20 and to correct several

editorial oversights previously evaluated. The proposed change has no

affect on any plant operating parameters. Consequently, the proposed

change does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. Will operation of the facility according to this proposed change

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

previously evaluated?

The proposed change is to bring the LACBWR Technical Specifications

into conformance with the revised 10 CFR Part 20 and to correct several

editorial oversights previously evaluated. The proposed change is

administrative in nature. Further, the proposed change does not result

in any physical alteration to any plant system, and does not result in

any change in the method by which any safety-related system performs

its function. Consequently, the 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 according to this proposed change

involve a significant reduction in a margin of safety?

The margin of safety is the difference between the value of a

critical design, operating, or post accident parameter, and the value

of the parameter which would produce unacceptable results. The proposed

change does not affect any hardware, has no effect on the current

operating methodologies or actions which govern plant performance, and

does not affect any accident analysis parameter. Consequently, the

proposed change does not involve a significant reduction in a margin of

safety.

The NRC staff has determined based on its own no significant

hazards consideration evaluation 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: La Crosse Public Library, 800

Main Street, La Crosse, Wisconsin 54601

Attorney for licensee: Fritz Schubert, Esquire, Dairyland Power

Cooperative, 2615 East Avenue South, La Crosse, Wisconsin 54601

NRC Branch Chief: John H. Austin

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: November 11, 1993

Description of amendment request: The proposed amendments would

consolidate the Quality Verification Department with the Nuclear

Generation Department and realign the Nuclear Safety Review Board such

that it reports to the Senior Vice-President of the Nuclear Generation

Department.

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 amendments do not involve a significant increase in the

probability or consequences of an accident previously evaluated.]

The proposed revisions to consolidate the Quality Verification

Department with the Nuclear Generation Department and realign the

NSRB [Nuclear Safety Review Board] such that it reports to the

Senior Nuclear Officer, change the reference from Semiannual to

Annual, change the reference from group to division, delete titles

of persons designated to approve modifications, clarify the

responsibilities of the Safety Assurance Manager, and delete the

requirement to perform an annual independent Fire Protection Audit

will not involve a significant increase in the probability or

consequences of an accident previously evaluated because the changes

do not have any impact upon the design or operation of any plant

systems or components.

[2. The amendments do not create the possibility of a new or

different kind of accident from any accident previously evaluated.]

The proposed revisions will not create the possibility of a new

or different kind of accident from any previously evaluated because

the changes are administrative in nature and operation of Catawba,

McGuire, and Oconee Nuclear Stations in accordance with these TS

[technical specifications] will not create any failure modes not

bounded by previously evaluated accidents.

[3. The amendments 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: 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, et al., Docket Nos. 50-413 and 50-414, Catawba

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

Date of amendment request: November 11, 1993

Description of amendment request: The proposed amendments would

consolidate the Quality Verification Department with the Nuclear

Generation Department and realign the Nuclear Safety Review Board such

that it reports to the Senior Vice-President of the Nuclear Generation

Department.

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 amendments do not involve a significant increase in the

probability or consequences of an accident previously evaluated.]

The proposed revisions to consolidate the Quality Verification

Department with the Nuclear Generation Department and realign the

NSRB [Nuclear Safety Review Board] such that it reports to the

Senior Nuclear Officer, change the reference from Semiannual to

Annual, change the reference from group to division, delete titles

of persons designated to approve modifications, clarify the

responsibilities of the Safety Assurance Manager, and delete the

requirement to perform an annual independent Fire Protection Audit

will not involve a significant increase in the probability or

consequences of an accident previously evaluated because the changes

do not have any impact upon the design or operation of any plant

systems or components.

[2. The amendments do not create the possibility of a new or

different kind of accident from any accident previously evaluated.]

The proposed revisions will not create the possibility of a new

or different kind of accident from any previously evaluated because

the changes are administrative in nature and operation of Catawba,

McGuire, and Oconee Nuclear Stations in accordance with these TS

[technical specifications] will not create any failure modes not

bounded by previously evaluated accidents.

[3. The amendments do not involve a significant reduction in a

margin of safety.]

The proposed revisions will not involve a reduction in a margin

of safety because they are administrative in nature.

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 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

Description of amendment request: The proposed amendments would

consolidate the Quality Verification Department with the Nuclear

Generation Department and realign the Nuclear Safety Review Board such

that it reports to the Senior Vice-President of the Nuclear Generation

Department. In addition, the requirement to conduct an annual

independent Fire Protection Audit is deleted.

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 amendments do not involve a significant increase in the

probability or consequences of an accident previously evaluated.]

The proposed revisions to consolidate the Quality Verification

Department with the Nuclear Generation Department and realign the

NSRB [Nuclear Safety Review Board] such that it reports to the

Senior Nuclear Officer, change the reference from Semiannual to

Annual, change the reference from group to division, delete titles

of persons designated to approve modifications, clarify the

responsibilities of the Safety Assurance Manager, and delete the

requirement to perform an annual independent Fire Protection Audit

will not involve a significant increase in the probability or

consequences of an accident previously evaluated because the changes

do not have any impact upon the design or operation of any plant

systems or components.

[2. The amendments do not create the possibility of a new or

different kind of accident from any accident previously evaluated.]

The proposed revisions will not create the possibility of a new

or different kind of accident from any previously evaluated because

the changes are administrative in nature and operation of Catawba,

McGuire, and Oconee Nuclear Stations in accordance with these TS

[technical specifications] will not create any failure modes not

bounded by previously evaluated accidents.

[3. The amendments do not involve a significant reduction in a

margin of safety.]

The proposed revisions will not involve a reduction in a margin

of safety because they are administrative in nature.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Loren R. Plisco, Acting

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: November 16, 1993

Description of amendment request: The proposed amendment would

revise the Technical Specifications to change the periodic test

schedule for containment Type A integrated leak rate tests (ILRTs) from

a set of three tests performed at approximately equal intervals during

each 10-year period, as specified in 10 CFR Part 50, Appendix J,

Section III.D, to one Type A test performed at 10-year intervals. The

change is being reviewed in conjunction with a proposed exemption to

Appendix J, as requested by the licensee in a letter dated November 16,

1993.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The Waterford 3 Type A test history provides substantial

justification for the proposed test schedule. Three type A tests

have been performed over an eight (8) year period with successful

results. The tests indicate that Waterford 3 has a low leakage

containment and that the leakage has never exceeded 24.6% of

La. [La is the maximum allowed leakage rate of air from

containment where containment is pressurized to Pa; for

Waterford 3 Pa is 44 psig. La for Waterford is 0.50

percent by weight of the containment air per 24 hours at Pa.]

There are no structural mechanisms which would adversely affect

the structural capability of the containment and that would be a

factor in extending the Type A test schedule to ten years. A risk

impact assessment was performed, and a determination was made that

there is no risk impact as a result of changing the Type A test

schedule. Therefore, the proposed change will not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

There are no design changes being made that would create a new

type of accident or malfunction. The proposed change will not alter

the plant or the manner in which it is operated. The change proposes

a change to the schedule for performing the periodic Type A test.

The purpose of the test is to provide periodic verification by test

of the leaktight integrity of the primary reactor containment, and

systems and components which penetrate containment. The tests assure

that leakage through containment and systems and components

penetrating containment will not exceed the allowable leakage rate

values associated with conditions resulting from an accident. The

change in schedule for performing the Type A test will not adversely

affect the containment integrity in the event of an accident.

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

new or different type of accident from any accident previously

evaluated.

The proposed change is a change to the schedule for performing

the periodic Type A tests and does not reduce the margin of safety

assumed in accident analysis for release of radioactive materials

from the containment atmosphere into the environment or any margin

of safety preserved by the Technical Specifications. The

methodology, acceptance criteria, and the technical specification

leakage limits for the performance of the Type A tests will not

change, and the Type A tests will be performed in accordance with

10CFR 50, Appendix J, and the Waterford 3 licensing basis.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: November 16, 1993

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to provide acceptable

conditions for operation when (1) the core operating limits supervisory

system (COLSS) is in service and neither control element assembly

calculator (CEAC) is operable and (2) the COLSS is out of service and

either or both CEACs are operable.

This proposed TS change modifies the departure from nucleate

boiling ratio (DNBR) margin, Limiting Condition for Operation (LCO)

3.2.4b and c, which limits the core power distribution to the initial

value assumed in the accident analyses. Operation within this LCO

either limits or prevents potential fuel cladding failures in the event

of a postulated accident and limits damage to the fuel cladding during

an accident by ensuring that the plant is operating within acceptable

conditions at the onset of a transient. The limiting safety system

settings and this LCO are based on the accident analysis, so that

specified acceptable fuel design limits (SAFDLs) are not exceeded as a

result of anticipated operational occurrences (AOOs) and the limits of

acceptable consequences are not exceeded for other postulated

accidents.

The COLSS and core protection calculators (CPCs) monitor the core

power distribution on line and are capable of verifying that the linear

heat rate (LHR) and DNBR do not exceed their limits. The COLSS performs

this function by continuously monitoring the core power distribution

and calculating core power operating limits corresponding to the

allowable peak LHR and DNBR. The CPCs perform this function by

continuously calculating an actual value of DNBR and LHR for comparison

with the respective trip setpoints. CEACs monitor CEA position. Should

a CEA deviate from its subgroup position, the CEACs will transmit an

appropriate ``penalty'' factor to the CPCs.

The COLSS is normally used to monitor DNBR margin. When at least

one CEAC is operable, TS 3.2.4a provides enough margin to DNB to

accommodate the limiting AOO without failing the fuel. When neither

CEAC is operable, the CPCs lack the CEA position information necessary

to ensure a reactor trip when necessary. In this case TS 3.2.4b

requires the COLSS calculated core power to be reduced to ensure that

the limiting AOO will not result in fuel failure. Currently, TS 3.2.4b

requires that the COLSS calculated power be maintained at 13% below the

COLSS calculated power operating limit to compensate for the potential

error in the CPC DNBR calculation. The proposed revision would increase

this required adjustment to 16%, which is more restrictive than the

present value.

In instances when the COLSS is out of service, but either or both

CEACs are operable, TS 3.2.4c states that the DNBR operating margin

shall be maintained by comparing the DNBR indicated on any operable CPC

channel with the allowable value from TS Figure 3.2-2. Whenever the

COLSS is out of service, the CPCs are used to perform the same

monitoring function. However, the extra conservatisms built into the

CPCs for transient protection are not all required when the CPCs are

being used for monitoring. In order not to affect the CPC transient

protection, these conservatisms are not taken from the CPC, but are

credited in the COLSS out-of-service limits in Figure 3.2-2. A

reevaluation of the limiting AOOs has verified that, by maintaining the

margin in the proposed Figure 3.2-2, sufficient margin exists to ensure

that the limiting Cycle 7 AOO will not result in fuel failure.

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:

For the case when neither CEAC is operable but COLSS is in

service, the CPCs assume a preset CEA configuration and can not

obtain the required CEA position information to ensure the SAFDL on

DNBR will not be violated during an AOO. Thus, as a result of

limiting AOOs for Cycle 7, Specification 3.2.4b requires that core

power be reduced to a value 16% less than the current COLSS

calculated power operating limit. This ensures the limiting AOO will

not result in a violation of SAFDLs. The proposed revision to Figure

3.2-2 accounts for the situation when COLSS is out of service but at

least one CEAC is operable. In this case, the Cycle 7 safety

analysis has shown that by maintaining the CPC calculated DNBR above

the value shown in the figure, the limiting AOO will not result in a

violation of the SAFDLs. Therefore, the proposed change will not

significantly increase the probability or consequences of any

accident previously evaluated.

The proposed changes are primarily a result of changes in Cycle

7 core parameters. These changes do not involve any change to any

equipment or manner in which the plant will be operated. These

changes further restrict the plant operation when either COLSS or

both CEACs are out of service. Therefore, the proposed change will

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

from any accident previously evaluated.

The intent of this Specification is to ensure that there is

always sufficient margin to DNB such that the CPCs can mitigate the

consequences of the most limiting AOO prior to a violation of the

SAFDLs. Generally, this margin is continuously monitored by COLSS;

however, if COLSS is out of service, but at least one CEAC is

operable, the limitation on CPC calculated DNBR (as a function of

ASI [axial shape index]) shown in Figure 3.2-2 represents a

conservative envelope of operating conditions consistent with the

Cycle 7 safety analysis assumptions. This band of operating

conditions has been analytically demonstrated to maintain an

acceptable minimum DNBR through all AOOs. On the other hand, for the

case when COLSS is in service, but neither CEAC is operable, the

proposed change will ensure that the limiting AOO will not result in

a violation of SAFDLs. Therefore, the proposed changes will not

result in a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile

Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: December 2, 1993

Description of amendment request: The purpose of the request is to

change the plant Technical Specifications (TS) to remove the limiting

conditions for operation and surveillance requirements for the chlorine

detection system. TMI-1 removed the gases Chlorination System for the

Circulating Water and River Water Systems. This modification eliminated

the need for a Chlorine Dectection System (CDS).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The TS requirements assured the operability of

the CDS in the event of an on-site chlorine release from a one ton

cylinder. These TS requirements reduced the probability and the

consequences of a radiological accident which may result from an

incapacitation of control room operatorsafter entry of chlorine into

the control room. With the removal and the restriction on delivery

of one ton chlorine cylinders, this postulated event is no longer

credible, and there is a decrease in the probability of a

radiological accident.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated. The TS

requirements associated with the CDS were for the on-site release of

chlorine from a one ton cylinder. These cylinders are removed and

prohibited from the TMI-1 site. These actions preclude a significant

on-site release of chlorine which could affect the control room

operators.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The purpose of the TS requirements was to maintain

operability of the CDS in the event of on-site release from a one

ton chlorine cylinder. Since chlorine cylinders greater than 150

pounds are prohibited on-site, the TS requirements for chlorine

detection are no longer required, and their removal will not reduce

the margin of safety.

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

this review, it appears that the three standards of 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, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

NRC Project Director: John F. Stolz

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

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

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: November 23, 1993

Description of amendment request: The licensee proposes to modify

the South Texas Project, Units 1 and 2, Technical Specification 3/

4.8.1.1, ``A.C. Sources,'' to modify the action statements and

surveillance requirements for testing of the standby diesel generator.

This amendment would incorporate the recommendations of NRC Generic

Letter (GL) 93-05, ``Line-Item Technical Specifications Improvements To

Reduce Surveillance Requirements For Testing During Power Operation,''

dated September 27, 1993.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change seeks to eliminate the unnecessary testing

of an operable Standby Diesel Generator (SDG). Technical

Specification (TS) 3.8.1.1 Actions a. and e. require all operable

SDGs be started as a demonstration of operability whenever one or

more of the offsite AC [alternating current] power sources is

declared inoperable. The inoperability of an offsite AC power source

has no effect on the reliability of a SDG. Deleting this requirement

does not affect the design or performance characteristics of the

SDGs. Therefore, the SDGs maintain their ability to perform their

design function.

TS 3.8.1.1 Actions b. and c. require all remaining operable SDGs

be started as a demonstration of operability whenever one of the SDG

is declared inoperable except for preplanned preventive maintenance

or testing. The proposed amendment would expand the testing

exclusion to include an inoperable support system and an

independently testable component in addition to preplanned

preventive maintenance and testing. The proposed amendment would

also eliminate the testing requirement of the remaining operable

SDGs, when a SDG is declared inoperable, unless there is cause to

believe a potential common mode failure exists for the remaining

SDGs. The normal TS surveillance testing schedule assures that

operable SDG(s) are capable of performing their intended safety

functions. A failure of one SDG does not reduce the reliability of

another, otherwise operable SDG. Deleting this requirement does not

affect the design or performance characteristics of the SDGs, once a

common mode failure has been dismissed. Therefore, the SDGs maintain

their ability to perform their design function.

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

or different kind of accident from any previously evaluated.

The elimination of these unnecessary tests does not affect the

design bases of the SDGs, or any of the accident evaluations

involving the SDGs. The SDGs are designed to provide electrical

power to the equipment important for safety during all modes and

plant conditions following a loss of offsite power. The test

schedule established in accordance with GL 84-15 [``Proposed Staff

Actions To Improve and Maintain Diesel Generator Reliability'']

assures that operable SDGs are capable of performing their intended

safety function. Therefore, this change does 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.

Since the proposed change does not affect the design bases,

accident analysis, reliability or capability of the SDGs to perform

their intended safety function, this change does not involve any

reduction in a margin to safety.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Project Director: Suzanne C. Black

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

Nuclear Plant, Unit No. 1, Berrien County, Michigan

Date of amendment request: December 15, 1993. This submittal

supersedes a previous submittal dated March 10, 1993.

Description of amendment request: The proposed amendment would

implement interim tube plugging criteria for the tube support plate

elevation outer diameter stress corrosion cracking for cycle 14.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of Donald C. Cook Nuclear Plant Unit 1 in

accordance with the proposed license amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Testing of model boiler specimens for free span tubing (no TSP

restraint) at room temperature conditions shows burst pressures in

excess of 5000 psi for indications of ODSCC with voltage

measurements as high as 19 volts. Burst testing performed on pulled

tubes from Cook Nuclear Plant Unit 1 with up to a 2.02 volt

indication shows measured burst in excess of 10,000 psi at room

temperature. Correcting for the effects of temperature on material

properties and minimum strength levels (as the burst testing was

done at room temperature), tube burst capability significantly

exceeds the RG 1.121 criterion requiring the maintenance of a margin

of 3 times normal operating pressure differential on tube burst. The

3 times normal operating pressure differential for the Cook Nuclear

Plant Unit 1 steam generators corresponds to 4275 psi. Based on the

existing data base, this criterion is satisfied with 7/8'' diameter

tubing with bobbin coil indications with signal amplitudes less than

4.9 volts, regardless of the indicated depth measurement. A 1.0 volt

plugging criteria compares favorably with the structural limit

considering the previously calculated growth rates for ODSCC within

the Cook Nuclear Plant Unit 1 steam generators. Considering a

voltage increase of 0.4 volts, and adding a 20% NDE uncertainty of

0.20 volts (90% Cumulative Probability) to the IPC of 1.0 volts

results in an EOC voltage of approximately 1.6 volts for Cycle 14

operation. A 3.3 volt safety margins implied (4.9 structural limit -

1.6 volt EOC - 3.3 volt margin). This EOC voltage compares favorably

with the Structural Limit of 4.9 volts.

For the voltage/burst correlation, the EOC structural limit is

supported by a voltage of 4.9 volts. A 3.1 volt BOC repair limit

confirms the structural limit when 40% growth and 20% uncertainty

are applied to the repair limit. This repair limit will be applied

for Cycle 14 IPC implementation to repair bobbin indication greater

than 3.1 volts independent of RPC confirmation of the indication.

The conservatism of this repair limit is shown by the EOC 12

(Summer 1992) eddy current data. The overall average voltage growth

was determined to be only 2.2%, with a 12% average voltage growth

for indications less than 0.75 volt BOC and a 1% average voltage

growth for indication >0.75 volt at the BOC. In addition, the Cycle

12 maximum observed voltage increase was found to be 0.49 volts, and

occurred in a tube initially 1.0 volt

bobbin indications with confirmed RPC calls, using the Cook Nuclear

Plant Unit 1 pulled tube destructive examination results, it is

reasonable that no indications of degradation greater than 40% to

56% deep with an ability to influence tube burst capability were

left in service. Since the majority of the EOC 14 indications at

Cook Nuclear Unit 1 are expected to be below this level, the

inclusion of all IPC intersections into the leakage calculation is

exceptionally conservative.

Therefore, as re-implementation of the 1.0 volt IPC during Cycle

14 does not adversely affect steam generator tube integrity and

results in acceptable dose consequences, the proposed amendment does

not result in any increase in the probability or consequences of an

accident previously evaluated within the Cook Nuclear Plant Unit 1

FSAR.

2. The proposed license amendment does not create the

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

previously evaluated.

Implementation of the proposed steam generator tube IPC does not

introduce any significant changes to the plant design basis. Use of

the criteria does not provide a mechanism which could result in a

tube rupture outside of the region of the TSP elevations; no ODSCC

is occurring outside the thickness of the TSPs. Neither a single or

multiple tube rupture event would be expected in a steam generator

in which the plugging criteria has been applied (during all plant

conditions).

Specifically, Cook Nuclear Plant will continue to implement a

maximum leakage rate limit of 150 gpd (0.1 gpm) per steam generator

to help preclude the potential for excessive leakage during all

plant conditions. The Cycle 14 Technical Specification limits on

primary to secondary leakage at operating conditions is a maximum of

0.4 gpm (600 gpd) for all steam generators, or, a maximum of 150 gpd

for any one steam generator. The RG 1.121 criterion for establishing

operational leakage rate limits that require plant shutdown are

based upon leaks-before-break consideration to detect a free span

crack before potential tube rupture. The 150 gpd limit should

provide for leakage detection and plant shutdown in the event of the

occurrence of an unexpected single crack resulting in leakage that

is associated with the longest permissible crack length. RG 1.121

acceptance criteria for establishing operating leakage limits are

based on leak-before-break considerations such that plant shutdown

is initiated if the leakage associated with the longest permissible

crack is exceeded. The longest permissible crack is the length that

provides a safety factor of 3 against bursting at normal operating

pressure differential. A voltage amplitude of 4.9 volts for typical

ODSCC corresponds to meeting this tube burst requirement at a lower

95% prediction limit on the burst correlation coupled with 95/95 LTL

material properties. Alternate crack morphologies can correspond to

4.9 volts so that a unique crack length is not defied by the burst

pressure versus voltage correlation. Consequently, typical burst

pressure versus through-wall crack length correlations are used

below to define the ``longest permissible crack'' for evaluating

operating leakage limits.

At current plant conditions, the single through-wall crack

lengths that result in tube burst at 3 times normal operating

pressure differential and SLB conditions are 0.44 inch and 0.84

inch, respectively. A leak rate of 150 gpd will provide for

detection of 0.42 inch long cracks at nominal leak rates and 0.61

inch long cracks at the lower 95% confidence level leak rates. Since

tube burst is precluded during normal operation due to the proximity

of the TSP to the tube and the potential for the crevice to become

uncovered during SLB conditions, the leakage from the maximum

permissible crack must preclude tube burst at SLB conditions. Thus,

the 150 gpd limit provides for plant shutdown prior to reaching

critical crack lengths for SLB conditions.

3. The proposed license amendment does not involve a significant

reduction in margin of safety.

The use of the voltage based bobbin probe interim TSP elevation

plugging criteria at Cook Nuclear Plant Unit 1 is demonstrated to

maintain steam generator tube integrity commensurate with the

criteria of Regulatory Guide 1.121. RG 1.121 describes a method

acceptable to the NRC staff for meeting GDCs 14, 15, 31, and 32 by

reducing the probability or the consequences of steam generator tube

rupture. This is accomplished by determining the limiting conditions

of degradation of steam generator tubing, as established by

inservice inspection, for which tubes with unacceptable cracking

should be removed from service. Upon implementation of the criteria,

even under the worst case conditions, the occurrence of ODSCC at the

TSP elevations is not expected to lead to a steam generator tube

rupture event during normal or faulted plant conditions. The EOC 14

distribution of crack indications at the TSP elevations will be

confirmed to result in acceptable primary to secondary leakage

during all plant conditions and that radiological consequences are

not adversely impacted.

In addressing the combined effects of LOCA + SSE on the steam

generator component (as required by GDC 2), it has been determined

that tube collapse may occur in the steam generators at some plants.

This is the case as the TSPs may become deformed as a result of

lateral loads at the wedge supports at the periphery of the plant

due to the combined effects of the LOCA rarefaction wave and SSE

loadings. Then, the resulting pressure differential on the deformed

tubes may cause some of the tubes to collapse.

There are two issues associated with steam generator tube

collapse. First, the collapse of steam generator tubing reduces the

RCS flow area through the tubes. The reduction in flow are increases

the resistance to flow of steam from the core during a LOCA which,

in turn, may potentially increase peak clad temperature (PCT).

Second, there is a potential that partial through-wall cracks in

tubes could progress to through-wall cracks during tube deformation

or collapse.

Consequently, since the leak-before-break methodology is

applicable to the Cook Nuclear Plant Unit 1 reactor coolant loop

piping, the probability of breaks in the primary loop piping is

sufficiently low that they need not be considered in the structural

design of the plant. The limiting LOCA event becomes either the

accumulator line brake or the pressurizer surge line break. LOCA

loads for the primary pipe breaks were used to bound the Cook

Nuclear Plant Unit 1 smaller breaks. The results of the analysis

using the larger break inputs show that the LOCA loads were found to

be of insufficient magnitude to result in steam generator tube

collapse or significant deformation.

Addressing RG 1.83 consideration, implementation of the bobbin

probe voltage based interim tube plugging criteria of 1.0 volt is

supplemented by the following: enhanced eddy current inspection

guidelines to proved consistency in voltage normalization, a 100%

eddy current inspection sample size at the TSP elevations, and RPC

inspection requirements as outlined in the technical specifications

and Appendix A ``NDE Data Acquisition and Analysis Guidelines''

(Attachment 6).

As noted previously, implementation of the TSP elevation

plugging criteria will decrease the number of tubes which must be

repaired. The installation of steam generator tube plugs reduce the

RCS flow margin. Thus, implementation of the alternate plugging

criteria will maintain the margin of flow that would otherwise be

reduced in the event of increased tube plugging.

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

amendment request does not result in a significant reduction in

margin with respect to plant safety as defined in the Final Safety

Analysis Report or any of the plant 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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Leaks Preston Palenske

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

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

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

NRC Project Director: A Randolph Blough, Acting

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,

Michigan

Date of amendment requests: November 15, 1993

Description of amendment requests: The proposed amendments delete

certain Limiting Conditions for Operation, Actions, and Surveillance

Requirements for Reactor Coolant System Pressure Isolation Valves in

the Technical Specifications. The Technical Specifications for these

Reactor Coolant System Pressure Isolation Valves were added by Order

dated April 20, 1981. This Order was prompted by concerns for an

interfacing system loss-of-coolant accident as identified in the

Reactor Safety Study (WASH-1400). The proposed Technical Specification

change, by inference, also requests rescission of the April 20, 1981

Order.

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:

Per 10 CFR 50.92, a proposed amendment to an operating license

will not involve a significant hazards consideration if the proposed

amendment satisfies the following three criteria:

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

consequences of an accident previously analyzed,

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

accident from an accident previously analyzed or evaluated, or

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

safety.

Criterion 1

The ISLOCA is not one of the accidents previously analyzed in

Chapter 14, Safety Analysis, of the Cook Nuclear Plant Updated Final

Safety Analysis Report. Chapter 14 analyzes the large break LOCA in

Section 14.3.1, and ``loss of reactor coolant from small ruptured

pipes or from cracks in large pipes which actuates the ECCS'', or

small break LOCA in Section 14.3.2. Therefore, deleting from the

Technical Specifications the Reactor Coolant System pressure

isolation valves in Table 3.4-0, will not increase the probability

or the consequences of the large break or the small break LOCAs

previously analyzed for the Cook Nuclear Plant.

Criterion 2

The Reactor Coolant System pressure isolation valves in Table

3.4-0 of the Technical Specifications were added because WASH-1400

identified the ISLOCA as a significant contributor to core damage

frequency. Deletion of the subject valves from the Technical

Specifications and reliance on the testing requirements mandated by

the In-Service Testing Program of ASME XI does not create the

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

break or the small break LOCAs previously analyzed for the Cook

Nuclear Plant.

Criterion 3

Deleting the Reactor Coolant System pressure isolation valves

from the testing requirements in Table 3.4-0 of the Technical

Specifications will result in these valves only being tested on a

refueling outage frequency as part of the ASME B&PV Code Section XI

IST Program. This somewhat reduced testing frequency will result in

a slight increase in the ISLOCA contribution to core damage

frequency of 5.4%, from lower 5.00E-08/reactor year to mid 5.00E-08/

reactor year. This insignificant increase will not affect the

overall core damage frequency of 6.26E-05/reactor year. Therefore,

it is concluded that the proposed deletion of the Reactor Coolant

System pressure isolation valves in Table 3.4-0 of the Technical

Specifications, as well as the proposed deletion of the portions of

the Technical Specifications that are affected by Table 3.4-0, will

not result in a significant reduction in the margin of safety that

exists at Cook Nuclear Plant to prevent an ISLOCA or to mitigate the

consequences of an ISLOCA.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

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

NRC Project Director: A. Randolph Blough, Acting

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay

Power Plant, Unit 3, Humboldt County, California

Date of application for amendment: October 8, 1993 (Reference LAR

93-02)

Brief description of amendment: This Licensee Amendment Request

(LAR) proposes to revise the Humboldt Bay Power Plant (HBPP), Unit 3,

Technical Specifications (TS) by deleting Figure II-2 in Section II,

``Site,'' by deleting the Restricted Area boundary line in Figure V-3,

Section V, ``Monitoring Systems,'' by incorporating a title change into

Section VII, ``Administrative Controls,'' and by revising Figure VII-2,

``Plant Staff Organization.'' The proposed changes are in response to

the revised 10 CFR Part 20 which becomes mandatory on January 1, 1994

(56 FR 23360). The specific TS changes proposed are as follows:

(1) Page v, Figures - delete reference to Figure II-2.

(2) Page II-1, Section II.B, Plant Areas - change ``is shown in

Figure II-2'' to ``shall be defined in plant procedures.''

(3) Page II-3, Section II - delete Figure II-2.

(4) Page V-14, Section V - delete the Restricted Area boundary line

from Figure V-3, ``HBPP Groundwater Monitoring Systems Wells,'' to be

consistent with item 3 above.

(5) Page VII-5, Section VII.C.2.e, Supervisor of Maintenance -

change the title from ``Supervisor of Maintenance'' to ``Maintenance

Planner.''

(6) Page VII-10, Section VII.D.1.b., Membership, List of minimum

membership - replace ``Supervisor of Maintenance'' with ``Maintenance

Planner.''

(7) Page VII-31, Section VII, Figure VII-2, Plant Staff

Organization - replace ``Maintenance Supervisor'' with ``Maintenance

Planner.''

(8) Page VII-31, Section VII, Figure VII-2, Plant Staff

Organization - both the Mechanical Foreman and the Instrument/

Electrical Foreman report directly to the Plant Manager, not to the

``Maintenance Planner,'' as previously shown.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

a. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

A change to the defined restricted area has no affect on any

plant operating parameters. Consequently, a change to the defined

restricted area will not affect the probability or consequences of

an accident occurring.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed revisions to the HBPP TS are administrative in

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

physical alteration to any plant system, and there would not be a

change in the method by which any safety-related system performs its

function.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

of safety?

The proposed revisions to the HBPP TS do not affect the margin

of safety of any accident analysis since they do not affect the

parameters for any accident analysis, and have no effect on the

current operating methodologies or actions which govern plant

performance.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Humboldt County Library, 636 F

Street, Eureka, California 95501

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

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

NRC Branch Chief: John H. Austin

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

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

Pennsylvania

Date of amendment request: October 29, 1993

Description of amendment request: The amendment would revise the

Limerick Generating Station, Units 1 and 2, Technical Specifications to

eliminate the Main Steam Line Radiation Monitoring System high

radiation trip function for initiating 1) an automatic reactor scram

and automatic closure of the Main Steam Line Isolation Valves, and 2)

automatic closure of the Main Steam Line drain valves, and Main Steam

and Reactor Water Sample line valves.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS changes involve eliminating the Main Steam Line

Radiation Monitoring (MSLRM) system high radiation trip function for

initiating an automatic reactor scram and automatic closure of the

Main Steam Line Isolation Valves (MSIVs), Main Steam line drain

valves, and Main Steam and Reactor Water sample line valves. The

proposed TS changes support installation of a plant modification to

defeat portions of MSLRM system high radiation trip function logic

circuitry in the Reactor Protection System (RPS) and Primary

Containment and Reactor Vessel Isolation Control System (PCRVICS).

Installation of this modification will not adversely impact the

operation of the RPS or PCRVICS with respect to performing its other

intended safety functions. The proposed TS changes will not affect

the operation of other plant systems or equipment important to

safety. The MSLRM system high radiation trip function for the

Mechanical Vacuum Pump (MVP) will be retained. The safety assessment

and justification for eliminating the MSLRM system high radiation

trip function for initiating an automatic reactor scram and

automatic closure of the MSIVs [are] based on General Electric's

(GE's) Topical Report NEDO-31400A, ``Safety Evaluation for

Eliminating the Boiling Water Reactor Main Steam Line Isolation

Valve Closure Function and Scram Function of the Main Steam Line

Radiation Monitor,'' and the applicability of this report to

Limerick Generating Station (LGS), Units 1 and 2. By letter dated

May 15, 1991, the NRC approved this topical report and indicated

that it was acceptable for licensees to reference this report as the

basis for requesting a TS change to eliminate the MSLRM system high

radiation trip functions as documented in the report and associated

NRC Safety Evaluation Report (SER).

The safety assessment provided in NEDO-31400A can also be

applied to eliminate the MSLRM system high radiation trip function

for initiating the automatic closure of the Main Steam line drain

valves although this aspect was not explicitly evaluated in NEDO-

31400A. The flow from these valves ultimately discharges to the main

condenser as do the MSIVs and therefore, any radioactive material

passing through these valves would be processed in the same fashion

as that passing through the MSIVs. The effects of eliminating the

MSLRM system high radiation trip function for initiating the closure

of the Main Steam and Reactor Water sample line valves is [are]

negligible. The sample lines are routed to a sample sink where inlet

valves installed on the sample lines are normally closed.

Additionally, downstream of the inlet valves are needle valves

designed to control and limit sample line flow. The sample sink is

enclosed, and air vented from its exhaust hood is passed through

filters prior to release to the environment. There is the potential

that a minimal amount of radioactive material could be released to

the environment if the sample sink inlet and needle valves failed to

properly function. This potential release has been evaluated and

determined to a small fraction of the dose limit requirements

specified in 10 CFR 100.

The MSLRM system high radiation trip was intended to function in

response to a Control Rod Drop Accident (CRDA), a Design Basis

Accident previously evaluated. Although the CRDA assumes MSIV

closure, no credit was taken for this in the CRDA analysis since it

postulates that the radioactive material calculated to be released

from the fuel is transported to the main condenser prior to the

MSIVs completely closing. Furthermore, the probability of a fuel

failure is independent of the operation of the MSLRM system.

The Steam Jet Air Ejectors (SJAEs) will continue to operate to

remove non-condensable gases from the main condenser for processing

by the Offgas Treatment system. The Offgas Treatment system will

continue to function as designed to reduce offgas radioactivity

levels prior to release to the environment. Eliminating the MSLRM

system high radiation isolation functions will improve operational

flexibility in that the main condenser will be available to aid in

decay heat removal. Elimination of the MSLRM system high radiation

trip functions in conjunction with proper operation of the Offgas

Treatment system will ensure that any radioactive material released

to the environment is a small fraction of 10 CFR 100 limits.

Therefore, the proposed TS changes associated with eliminating

the MSLRM system high radiation trip function for initiating an

automatic reactor scram and automatic closure of the MSIVs, Main

Steam line drain valves, and Main Steam and Reactor Water sample

line valves do not involve an increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed TS changes involve eliminating the MSLRM system

high radiation trip function for initiating an automatic reactor

scram and automatic closure of the MSIVs, Main Steam Line drain

valves, and Main Steam and Reactor Water sample line valves. The

proposed TS changes will not affect the operation of other plant

systems or equipment important to safety. The associated plant

modification simply defeats the MSLRM system high radiation trip

function logic circuitry in the RPS and PCRVICS. The RPS and PCRVICS

will continue to respond in performing its other design intended

safety functions. The MSLRM system high radiation trip function for

the MVP will be retained. The proposed TS changes do not involve any

plant hardware changes that could introduce any new failure modes or

effects. The MSLRM system radiation monitors will remain active to

initiate Main Control Room (MCR) annunciation alarms. Plant

procedures will be in place to implement the appropriate mitigative

measures in response to a MSLRM system high radiation alarm signal.

The SJAEs will continue to operate to remove non-condensable

gases from the main condenser for processing by the Offgas Treatment

system. The Offgas Treatment system will continue to function as

designed to reduce offgas radioactivity levels prior to release to

the environment.

Since the Design Basis Accident analysis (i.e., CRDA) does not

credit the MSLRM system high radiation trip function for reducing

the radiological consequences of the postulated accident, the

proposed TS changes have effectively been evaluated and are included

in the existing analysis. That is, the CRDA analysis already assumes

that the radioactive material released from the failed fuel is

immediately transported to the main condenser prior to the MSIVs

completely closing.

The safety assessment and assumptions documents in GE Topical

Report NEDO-31400A provide the basis for eliminating the MSLRM

system high radiation trip function for initiating an automatic

reactor scram and automatic closure of the MSIVs. The safety

assessment provided in NEDO-31400A can also be applied to eliminate

the MSLRM system high radiation trip function for initiating the

closure of the Main Steam Line drain valves, since any radioactive

material passing through these valves would be processed in the same

fashion as that passing through the MSIVs. Eliminating the MSLRM

system high radiation trip function for initiating the closure of

the Main Steam and Reactor Water sample line valves will have a

negligible impact. The sample lines are routed to a sample sink

where inlet valves installed on the sample lines are normally

closed. Downstream of the inlet valves are needle valves designed to

control and limit sample line flow. The sample sink is located in

the Reactor Enclosure and is enclosed, and air vented from its

exhaust hood is passed through filters prior to release to the

environment. The Reactor Enclosure ventilation duct radiation

monitor samples air from the sample sink hood exhaust, and will

isolate the Reactor Enclosure ventilation system if the radiation

levels exceed the monitor's setpoint. There is the potential that a

minimal amount of radioactive material could be released to the

environment through this flowpath if the sample sink inlet and

needle valves failed to properly function. This potential release

has been evaluated and determined to a small fraction of the dose

limit requirements specified in 10CFR100.

Therefore, the proposed TS changes do not create the possibility

of a new or different kind of accident previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed TS changes to eliminate the MSLRM system high

radiation trip function for initiating an automatic reactor scram

and automatic closure of the MSIVs, Main Steam line drains valves,

and Main Steam and Reactor Water sample line valves do not change

the conclusion reached in the LGS Updated Final Safety Analysis

Report (UFSAR) that the calculated radiological consequences of the

bounding Design Basis Accident (i.e., CRDA) will not exceed the dose

limit requirements established by 10 CFR 100. The proposed TS

changes will improve the overall reliability of the plant when

compared to the existing system lineup configuration, since it will

reduce the potential of an unnecessary plant transient occurring as

a result of an inadvertent MSIV closure.

A reliability assessment analysis was performed to evaluate the

effects of eliminating the MSLRM system high radiation reactor scram

function on reactivity control failure frequency and core damage

frequency in GE Topical Report NEDO-31400A. This analysis indicated

that there is a negligible increase in reactivity control frequency

with the elimination of the MSLRM trip function. However, this

increase is compensated for by the reduction in transient initiating

events (i.e., inadvertent reactor scrams). This reduction in

transient initiating events represents a reduction in core damage

frequency and thus, results in a net improvement in safety.

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

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Larry E. Nicholson, Acting

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

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

Pennsylvania

Date of amendment request: November 30, 1993

Description of amendment request: The amendment would extend the

surveillance interval of the primary containment drywell-to-suppression

chamber bypass leak test from the current 18-month interval as required

by Technical Specification (TS) Surveillance Requirement 4.6.2.1.d to a

40 +/- 10-month interval. This change would allow the drywell-to-

suppression chamber bypass test to coincide with the 10 CFR 50,

Appendix J, Type A test (i.e., Containment Integrated Leakage Rate Test

(CILRT)) interval.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The failure effects that are potentially created by the proposed

Technical Specifications (TS) changes have been considered. The

accident which is potentially negatively impacted by the proposed TS

changes are any Loss of Coolant Accident (LOCA) inside primary

containment with or without offsite power available.

The proposed TS changes increase the surveillance interval of

the drywell-to-suppression chamber bypass leak test required by TS

Section 4.6.2.1.d, and will require that an additional test be

performed on the downcomer vacuum breakers assemblies. The primary

containment structure and associated equipment are not considered to

be accident initiators, they act to mitigate the consequences of an

accident. There are no physical or operational changes being made as

a result of these proposed changes. Therefore, the probability of

occurrence of an accident previously evaluated is not increased.

There is a potential increased risk that an increase in the

bypass leakage may go undetected for the duration of the proposed

extension of the interval between the performance of the drywell-to-

suppression chamber bypass leak test. However, as discussed below,

the increased risk is considered to be negligible due to the design

of the diaphragm structure and past test data. Therefore, we have

concluded that the probability of bypass leakage exceeding the

allowed value is not increased as a result of the proposed TS

changes.

The proposed TS changes will extend the surveillance interval

for the drywell-to-suppression chamber bypass leak test from 18

months to 40 +/- 10 months. These proposed changes would allow this

test to be performed at the same interval as the 10CFR50, Appendix

J, Type A test (i.e., Containment Integrated Leakage Rate Test

(CILRT)). In addition, the proposed changes will add an additional

surveillance requirement to be performed on the vacuum breaker

assemblies during refueling outages when the drywell-to-suppression

chamber bypass leak test is not required to be performed. The

proposed TS changes do not increase the consequences of an accident

previously evaluated. This is based on the evaluation summarized

below that demonstrates that the overall impact, if any, on the

plant containment integrity is negligible. Furthermore, the

performance history for the previous LGS bypass leak tests does not

indicate any time based failures. The proposed TS changes also

include a change to the frequency of testing, if two consecutive

tests fail, from once every nine (9) months to once every 24 months

in order to coincide with the 24 month refueling cycle. This change

has no impact on the consequences of an accident based on

maintaining the original requirement to increase the frequency of

testing if two consecutive bypass leak tests fail, and maintaining a

TS requirement for the NRC to review the schedule for subsequent

tests.

During a LOCA inside containment, potential leak paths between

the drywell and suppression chamber airspace could result in

excessive containment pressures, since the steam flow into the

airspace would bypass the heat sink capabilities of the suppression

pool. The containment pressure response to the postulated bypass

leakage can be mitigated by manually actuating the suppression

chamber sprays. Accordingly, since the sprays are manually actuated,

an analysis was performed to show that the operator has sufficient

time to initiate the sprays prior to exceeding the containment

design pressure. This analysis is described in section 6.2.1.1.5 of

the LGS Updated Final Safety Analysis Report (UFSAR). The analysis

is based on a small break LOCA inside containment with a

differential pressure between the drywell-to-suppression chamber

equal to the static pressure due to downcomer submergence. The

analysis concludes that the containment design pressure of 55 psig

will be reached in over 30 minutes from the onset of a small break

LOCA assuming a drywell-to-suppression chamber bypass flow area

(i.e., A/square root of k) equal to 7.20 in2 without operator

intervention.

TS Limiting Condition for Operation 3.6.2.1.b conservatively

specifies a maximum allowable bypass area of 10 % of the design

value of 7.20 in2. This TS limit provides an additional safety

factor of 10 above the conservatism taken in the steam bypass

analysis (i.e., 0.720 in2). The drywell-to-suppression chamber

bypass leak test required by TS Surveillance Requirement 4.6.2.1.d

verifies that the actual bypass flow area is less than or equal to

the TS limit of 0.720 in2. The bypass leakage test ensures that

degradation in the measured bypass area is identified and corrected

to ensure containment integrity during LOCA events.

The potential bypass leakage paths can be divided into two

categories as described below.

1) Leakage pathways other than those associated with the

drywell-to-suppression chamber vacuum breaker assemblies such as

diaphragm floor penetrations (i.e., downcomer and Safety/Relief

Valve (SRV) discharge line penetrations), cracks in the diaphragm

floor and/or liner plate, and cracks in the downcomers and SRV

discharge lines that pass through the suppression chamber airspace.

2) The four sets of drywell-to-suppression chamber vacuum

breaker assemblies.

All other potential bypass leakage pathways have at least two

isolation valves in the potential leakage path. These valves are

high quality leak-tight containment isolation valves that are

normally closed and receive an isolation signal to close. All Air

Operated Valves (AOVs) in these paths fail closed.

Several plant design features and the bypass leak test data

measured to date confirm that the leakage from other than the vacuum

breaker assemblies is negligible and indicates that this leakage

will continue to be negligible for the proposed increased duration

between tests. All pressure boundary penetrations between the

drywell and the suppression chamber are welded except the vacuum

breaker valves and the blind flanges closing 10 spare nozzles in the

downcomers. All pressure boundary penetrations between the drywell-

to-suppression chamber have been fabricated, erected, and inspected

in accordance with the American Society of Mechanical Engineers

(ASME) Code, Section III, Subsection NC, 1971 Edition, with the

exception of the tees supporting the vacuum breakers.

The downcomer and SRV discharge lines penetrate through the

diaphragm slab and terminate in the suppression pool. A steel ring

plate is welded to the outside of the downcomers. The downcomer/ring

plate assemblies are embedded in the diaphragm slab with the top

surface of the ring plate flush with the drywell side of the

diaphragm slab. All connections are welded to form a continuous

steel membrane between the liner plate and downcomer penetrations.

The SRV discharge lines are routed through welded flued heads at the

diaphragm floor. The flued head design and construction are similar

to the downcomer penetrations and also provide a continuous steel

barrier. The downcomer and SRV discharge lines are designed and

constructed to safety-related requirements. In addition, they are

designed for all postulated loading conditions, including seismic,

hydrodynamic, pressure, and temperature loads. The conservative

design requirements ensure that the SRV discharge and the downcomer

lines will not contribute to bypass leakage.

The diaphragm floor is a reinforced concrete slab approximately

3.5 feet thick. The drywell side surface of the diaphragm slab is

capped with a 1/4 inch thick carbon steel liner plate. The liner

plate and diaphragm slab provide a barrier against the potential for

bypass leakage through the diaphragm floor. The structural integrity

of the diaphragm floor and penetrations was demonstrated during the

pre-operational test program. The drywell was pressurized to a

drywell-to-suppression chamber differential pressure of above 30

psid, which envelopes the maximum drywell-to-suppression chamber

differential pressure postulated to occur during LOCA conditions.

There have been six Unit 1 and three Unit 2 bypass leak tests

performed in accordance with TS Surveillance Requirement 4.6.2.1.d.

These tests were conducted at a drywell-to-suppression chamber

differential pressure of at least 4.0 psid. The measured leakage

area includes leakage from both the vacuum breakers and sources

other than vacuum breakers.

In all cases, the measured leakage is significantly less than

the TS and design values. The maximum measured leakage areas are

0.0400 in2 and 0.0111 in2 for Unit 1 and Unit 2,

respectively; or 5.56% and 1.55 %, respectively, of the TS limit.

The average values are 0.0180 in2 for Unit 1 and 0.0107

in2 for Unit 2; or 2.5% and 1.49%, respectively, of the TS

limit of 0.720 in2. The minimum measured leakage areas are 0.0

in2 and 0.0100 in2 for Unit 1 and Unit 2, respectively, or

0% and 1.3 %, respectively, of the TS limit. Clearly, the test data

confirm that the bypass leakage measured to date at LGS has been

negligible.

In addition, we have obtained bypass leakage data from the

Pennsylvania Power and Light Company, Susquehanna Steam Electric

Station (SSES), Units 1 and 2, which also has Mark II containments

with the Anderson Greenwood vacuum breakers (i.e., the same

manufacturer as the vacuum breakers installed in the LGS, Unit 1 and

Unit 2 containments) and therefore the data is applicable to LGS.

The maximum bypass leakage area for the SSES Unit 1 containment was

0.037 in2, and 0.009 in2 for the SSES Unit 2 containment,

or 4.81% and 1.17%, respectively, of the SSES TS limit. Approval for

a similar TS change for SSES, Units 1 and 2 was issued by the NRC by

letter dated August 11, 1993.

The remaining and most likely source of potential bypass leakage

is the four sets of drywell-to-suppression chamber vacuum breakers.

Each set consists of two vacuum breakers in series, flange mounted

to a tee off the downcomers in the suppression chamber airspace. The

drywell-to-suppression chamber bypass leak test is currently

required by TS Surveillance requirement 4.6.2.1.d to be completed

during each refueling outage and the results are used to verify that

the total bypass area, including that due to the vacuum breakers,

meets the TS limit. If maintenance has been performed on the vacuum

breakers, this test also serves as a post-maintenance vacuum

breakers leakage area test.

The proposed TS changes decrease the frequency of the drywell-

to-suppression chamber bypass leak test. The drywell-to-suppression

chamber bypass leak test data obtained following vacuum breakers

maintenance cannot be utilized to determine vacuum breakers leakage

reliability over the duration of the proposed test interval

extension. To address this concern and collect additional vacuum

breakers leakage data, the proposed TS changes include an additional

requirement to perform a vacuum breaker leakage test as described

below.

The leakage test will be conducted on each set of vacuum

breakers (i.e., four vacuum breakers sets per unit) during each

refueling outage when the drywell-to-suppression chamber bypass leak

test would not be required to be performed. If maintenance is

performed on the vacuum breaker assemblies, this additional test

will be performed post-maintenance to verify that the leakage is

acceptable. This test will be conducted at a drywell-to-suppression

chamber differential pressure of 4.0 psid (i.e., the same as

differential pressure required for the drywell-to-suppression

chamber bypass leak test) by either pressurizing the drywell side of

the vacuum breakers or inducing a vacuum on the suppression chamber

side of the vacuum breakers. The acceptance criteria for the vacuum

breaker leakage tests will be as follows. The total vacuum breaker

leakage areas for all four sets of vacuum breakers will be less than

or equal to 24% of the TS limit (i.e., 0.24 x 0.720 in2 = 0.173

in2). This proposed acceptable vacuum breaker leakage area

provides a 76% margin to the TS limit to account for the leakage

paths other than the vacuum breakers. As described above, previous

bypass leakage testing measured a maximum bypass leakage area of

5.56% of the TS limit. The 76% margin is sufficiently large to

accommodate the other expected leakage sources. In addition, each

set of vacuum breakers will be limited to a leakage area twice the

assumed leakage from a single vacuum breaker set, assuming the

leakage area is evenly distributed among the four sets of vacuum

breakers (i.e., four sets equate to 24% of the TS Limit where each

set is 6% and twice this total is 12% of the TS Limit). This allows

a leakage of less than or equal to 0.0865 in2 (i.e., (0.173

in2 divided by 4 sets of vacuum breakers) x (a factor of 2

times the acceptable total) = 0.0865 in2) for an individual set

of vacuum breakers. This criterion is stipulated to identify

individual sets of vacuum breakers with higher leakage area.

The drywell-to-suppression chamber bypass leak test data

obtained during previous testing at LGS demonstrates conformance by

a large margin compared to the TS and design leakage requirements.

The test data indicates that there is negligible risk that the

bypass leakage will change adversely in future years. Furthermore,

the proposed test frequency is judged to be acceptable based on the

risk of the leakage sources other than the vacuum breakers being

essentially equivalent to that of the rest of the primary

containment structure, which is leak tested (i.e., CILRT) every 40

+/- 10 months as required by TS Surveillance Requirement 4.6.1.2.a.

A bypass leak test will be developed and conducted to verify

acceptable vacuum breaker bypass leakage areas for those outages

when the bypass leak test will not be required to be performed. The

proposed vacuum breaker leakage test with stringent acceptance

criteria, combined with other negligible leakage areas, provide an

acceptable level of assurance that the bypass leakage can be

measured and an adverse condition can be detected and corrected such

that the existing level of confidence that the primary containment

will function as required during a LOCA is maintained.

Therefore, the proposed TS changes will not involve an increase

in the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed TS changes involve the drywell-to-suppression

chamber bypass leak test frequency. There are no physical or

operational changes as a result of these proposed changes. These

proposed changes include the requirement to perform an additional

surveillance test on the vacuum breaker assemblies, applying a

differential pressure of 4.0 psid which is the same differential

pressure as currently required by TS for the drywell-to-suppression

chamber bypass leak test. This required test will ensure that

acceptable vacuum breaker leakage is maintained during those

intervals when the drywell-to-suppression chamber bypass leak test

is not required to be performed. Furthermore, the affected structure

(i.e., primary containment) acts as an accident mitigator and not as

an accident initiator. Accordingly, the possibility of a different

type of malfunction of equipment or the possibility of an accident

of a different type is not introduced.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The drywell-to-suppression chamber bypass leak test data

obtained during previous testing at LGS demonstrates conformance by

a large margin to the TS and design leakage requirements. The test

data indicate that there is negligible risk that the bypass leakage

will change adversely in future years. Furthermore, the proposed

test frequency is judged to be acceptable based on the risk of

sources of leakage other than the vacuum breakers being essentially

equivalent to that of the rest of the primary containment structure,

which is tested every 40 +/- 10 months. A bypass leak test will be

developed and conducted to verify acceptable vacuum breaker bypass

leakage areas for those outages when the bypass leak test will not

be required to be performed. The proposed vacuum breaker leakage

test with stringent acceptance criteria, combined with the other

negligible potential leakage areas, provide an acceptable level of

assurance that the bypass leakage can be measured and an adverse

condition can be detected and corrected such that the existing

levels of confidence that the primary containment will function as

required during a LOCA is maintained.

Therefore, the consequences of an accident are not impacted by

this change and containment integrity during a LOCA will be

maintained.

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

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Larry E. Nicholson, Acting

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: November 17, 1993

Description of amendment request: The proposed Technical

Specification (TS) changes to Surveillance Requirements would eliminate

unnecessary emergency diesel generator (EDG) testing when a diesel

generator or an offsite power source becomes inoperable. The proposed

change would reduce the stresses on the diesel generators caused by

unnecessary testing.

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 implementation of the proposed TS change, which

would delete the requirement to demonstrate the operability of an

otherwise operable EDG once the potential for a common cause failure

has been dismissed, does not affect the design or performance

characteristics of an EDG. Similarly, deleting the requirement to

demonstrate the operability of EDGs when an offsite power source is

inoperable does not affect the design or performance characteristics

of an EDG. Therefore, the EDGs will maintain their ability to

perform their design function. The EDGs are not assumed to be an

initiator of any analyzed event. The role of the EDGs is the

mitigation of accident consequences. Therefore, this proposed TS

change does not increase the probability of an accident previously

evaluated.

The consequences of an accident previously evaluated could be

affected by the proposed TS change. As described above,

implementation of the proposed change will result in the EDGs

maintaining their ability to perform their design function.

Excessive testing of EDGs can cause reduced reliability. Precluding

unnecessary testing of operable EDGs will improve EDG reliability

and thereby have an overall positive affect on plant safety.

Therefore, this proposed TS change does not increase the

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated

because implementation of the proposed TS change will not involve

physical changes to plant systems, structures, or components (SSC).

The design or performance characteristics of the EDG will not be

affected by the proposed change. The proposed change does not

introduce any new modes of plant operation or make any changes to

system setpoints which would initiate a new or different kind of

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

accident from any accident previously evaluated is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety because the proposed TS change does not affect

the design or performance of any EDG. The change will increase EDG

reliability by reducing the stresses on the EDG from unnecessary

testing. This will result in an overall increase in plant safety.

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 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.

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: November 19, 1993

Description of amendment request: The proposed change would

eliminate the listing of specific position titles for the Plant

Operations Review Committee (PORC) composition in favor of allowing the

Plant Manager to appoint PORC members. This would eliminate the need to

change the Technical Specifications (TSs) in the future whenever a

position title is changed.

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 TS change is administrative in

nature. The PORC member titles will be removed from the TS to

facilitate not requiring that a TS change be submitted for NRC

approval when position titles change. PORC member qualifications

will continue to be consistent with those required for the Facility

Staff and meet or exceed Sections 4.2, 4.4, or 4.6 of ANSI N18.1-

1971. The proposed change ensures that PORC will continue to be

comprised of personnel involved in daily plant activities who are

experienced individuals with varied expertise. By maintaining the

qualification requirements for members of PORC who represent various

areas of expertise, PORC will continue to fulfill its requirements

specified in TS Section 6.5.1.6. The proposed change does not

involve any physical changes to plant systems, structures, or

components (SSC), or the manner in which these SSC are operated,

maintained, modified, tested, or inspected. Therefore, the proposed

TS change does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated

because implementation of the proposed TS change will not involve

physical changes to plant SSC or the manner in which these SSC are

operated, maintained, modified, tested or inspected. The proposed

change does not introduce any new modes of plant operation or make

any changes to system setpoints which would initiate a new or

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

different kind of accident from any accident previously evaluated is

not created.

3. The proposed change does not involve a significant reduction

in a margin of safety because the proposed TS change is

administrative in nature by providing internal flexibility in

changing organizational titles and does not reduce the PORC function

or responsibilities. PORC will continue to be filled by

appropriately qualified personnel who have a variety of expertise.

The change does not affect the plant material condition, operation,

or accident analyses. 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 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.

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

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama

Date of amendments request: November 24, 1993

Description of amendments request: The proposed changes to the

Technical Specification will relocate the reactor trip system and

engineered safety feature actuation system response time limits from

the TS to the 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 changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes are administrative in nature and do not involve

any change to the configuration or method of operation of any plant

equipment used to mitigate the consequences of an accident. Also,

the proposed changes do not alter the conditions or assumptions in

any of the FSAR accident analyses. Since the FSAR accident analyses

remain bounding, the radiological consequences previously evaluated

are not adversely affected by the proposed changes. Therefore, the

proposed changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated. The proposed changes are administrative in nature and do

not involve any change to the configuration or method of operation

of any plant equipment used to mitigate the consequences of an

accident.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

changes. Therefore, the proposed changes do not create the

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

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety. The proposed changes are administrative in

nature and will continue to ensure that the response times for the

RTS and ESFAS instrumentation do not exceed the limits assumed in

the accident analyses. As a result of the proposed changes, response

time limits for the RTS and ESFAS will be administratively

controlled in accordance with the provisions of 10 CFR 50.59, thus

eliminating an unnecessary burden of governmental regulation without

reducing protection for public health and safety. Therefore, the

proposed changes do not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Attorney for licensee: James H. Miller, III, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201

NRC Project Director: S. Singh Bajwa

Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and

50-296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3,

Limestone County, Alabama

Date of amendment request: September 30, 1993 (TS 337)

Description of amendment request: The proposed amendments provide

an administrative vehicle for modifying a condition of the facility

operating license for each of the BFN units. The condition requires the

licensee to implement and maintain in effect all provisions of the

``Fire Protection Program (FPP)'' and lists the U.S. Nuclear Regulatory

Commission (NRC) staff safety evaluations (SE) approving the FPP. If

the staff approves of a revision currently under review to an element

of the FPP, the ``Appendix R Safe Shutdown Program (SSP)'', the

proposed amendments would add the staff SE documenting approval of the

revised SSP to the above listing of SEs in each facility operating

license. The current SSP is directed toward the safe shutdown of only

one operating plant (Unit 2). The revised SSP would be directed toward

the safe shutdown of two operating plants (Units 2 and 3).

Additionally, the proposed amendments add the definition of the SSP

to Section 1.0 of the Unit 3 Technical Specifications (TS).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

This proposed change is administrative in nature. The proposed

change is being made to revise the license condition to reflect a

combined Unit 2 and 3 Appendix R Safe Shutdown Program following NRC

approval. Compliance with the applicable Appendix R requirements is

ensured through implementation of the Fire Protection Program and

the Appendix R Safe Shutdown Program. The change does not affect any

design bases accident or the ability of any safe shutdown equipment

to perform its function. Also, there are no physical modifications

required to implement this TS change. Therefore, these proposed

administrative changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change is administrative in nature. The proposed

change is being made to revise the license condition to reflect a

combined Unit 2 and 3 Safe Shutdown Program following NRC approval.

Compliance with the applicable Appendix R requirements is ensured

through implementation of the Fire Protection Program and Appendix R

Safe Shutdown Program. This change does not affect any design basis

accident or the ability of any safe shutdown equipment to perform

its function. Also, there are no physical modifications required to

implement this TS change. Therefore, these proposed administrative

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

accident from an accident previously evaluated.

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

margin of safety.

The proposed changes are administrative in nature. Compliance

with the applicable Appendix R requirements is ensured through the

implementation of the Fire Protection Program and Appendix R Safe

Shutdown Program. The proposed change does not affect any design

basis accident and does not reduce or adversely affect the

capability to achieve and maintain safe shutdown in the event of a

fire. Furthermore, no reductions to the requirements for equipment

operability, surveillance requirements or setpoints are being made

which could result in reduction in the margin of safety. Therefore,

these proposed administrative changes will not result in a reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Mr. Frederick J. Hebdon

Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and

50-296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3,

Limestone County, Alabama

Date of amendment request: October 12, 1993 (TS 320)

Description of amendment request: The proposed amendment would

delete reference in the BFN Unit 3 Technical Specifications (TS) to the

Reactor Water Cleanup (RWCU) system floor drain high temperature

switches and the RWCU system space high temperature switches. The

piping configuration for the Unit 3 RWCU system has been modified, and

the licensee contends that its revised High Energy Line Break (HELB)

analysis has demonstrated that these switches are no longer required.

Instead, to initiate RWCU system isolation, the HELB analysis has

indicated the need for temperature switches in the main steam vault,

the heat exchanger room, and the RWCU pipe trench. The proposed

amendment therefore would add temperature switches to the Unit 3 TS for

these areas and modify the TS Bases section accordingly. The proposed

amendment also adds clarifying remarks to Tables 3.2.A and 3.2.B of the

TS for each of the BFN units. The proposed remarks list the actuation

signals for the various primary containment valve group isolations.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

An analysis of HELBs in the Unit 3 reactor building identified

certain RWCU pipe breaks which could not be automatically detected

and isolated in a reasonable time frame. To resolve this issue, a

design change is being performed to remove from service the existing

non-environmentally qualified temperature switches used to detect

RWCU line breaks and replace them with environmentally qualified

RTDs [resistance temperature detectors] and IEEE [Institute of

Electrical and Electronics Engineers] Class 1E qualified ATUs

[analog trip units] located to detect and isolate the critical RWCU

pipe breaks. This TS amendment adds the new ATUs [sic] function to

Tables 3.2.A and 4.2.A. Note 14 is deleted from Table 3.2.A since it

only applies to the temperature switches being removed from the

table.

The safety function of the RTD/ATU temperature loops is to

provide an isolation signal to close the RWCU suction line isolation

valves (FCV-69-001 and FCV-69-002) and RWCU return line valve (FCV-

69-012) on a high area temperature. This ensures RWCU pipe breaks

are isolated. No other RWCU safety functions are affected by the

change.

The new RTD/ATU temperature loops were chosen to decrease the

time required to initiate closure of the RWCU valves. This improves

the detection/isolation of RWCU breaks and helps to limit the

reactor coolant lost, helps ensure core cooling, and helps ensure

that environmental conditions inside the reactor building are

maintained within the required limits.

Components added by this change are qualified for the

environment in which they will operate. This ensures that the system

will perform its function in a post accident environment. No

additional paths for the release of radiation or contamination are

created. The failure modes of the RTDs and ATUs are such that any

single failure will result in a gross failure alarm and/or a channel

trip. Because of the redundancy, separations, and logic designed

into the system, a single failure of any part of the system will not

prevent isolation of the primary containment isolation valves and

spurious operation is minimized. The RTDs will be located and the

instrument setpoints will be set to preclude spurious trips due to

ambient temperatures including localized hot areas while assuring a

timely trip due to a pipe break. Therefore, the proposed amendment

does not involve a significant increase in the probability or

consequences of any accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

This change is being made to improve the RWCU leak detection/

isolation function of the RWCU Primary Containment Isolation System

(PCIS). The PCIS will perform its intended safety function in the

same manner as the previous installation. There is no affect [sic]

on the function or operation of any other plant system.

Failure of the RTD/ATU temperature loops would be no different

than failure of existing temperature switches. Since environmental

qualification requirements, divisional separation, single failure

requirements and one-out-of-two taken twice logic requirements are

maintained, the possibility of a RWCU isolation failure on a RWCU

line break or of a spurious isolation is no more likely after the

change than before.

In the existing design, logic relays are powered from RPS Bus A

or B. The new design also uses RPS Bus A or B to feed the ATUs.

Therefore, the consequence of a power failure is unchanged from the

present design. The seismic qualification and proper circuit

coordination of the installation is maintained. The system functions

and operates in the same manner as previously evaluated in the

Safety Analysis Report. No new system interactions other than

additional RTDs located in the main steam valve vault to input into

the PCIS logic for isolation of the RWCU have been introduced by

this activity. Therefore, the proposed amendment does not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The margin of safety will be enhanced by installing instruments

that provide quicker response to a temperature rise indicative of a

pipe break. Calculations have been performed to determine the

analytical limits for the RTD/ATU temperature loops in each of the

monitored areas and to determine the setpoints for the ATUs in each

area. The setpoints are set above the maximum expected room

temperatures to avoid spurious actuations due to ambient conditions

and below the analytical limits to ensure timely detection of a pipe

break. This type of design utilizing ATUs has been analyzed by the

NRC [U.S. Nuclear Regulatory Commission staff] (NEDO-21617, Analog

Transmitter/Trip Unit System for Engineered Safeguard Sensor Trip

Input) and has been found to be generically acceptable at BWR

facilities. Therefore, the proposed amendment does not involve a

significant 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: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Mr. Frederick J. Hebdon

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 10, 1993

Description of amendment request: The proposed change would change

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

Units No. 1 and No. 2 (NA-1&2).

Specifically, the proposed changes would modify the surveillance

frequency of the Auxiliary Feedwater System pumps from monthly to

quarterly in accordance with the guidance provided in Generic Letter

93-05, ``Line-Item Technical Specifications Improvements to Reduce

Surveillance Requirements for Testing During Power Operation,'' dated

September 27, 1993.

The NRC has completed a comprehensive examination of surveillance

requirements in TS that require testing at power. The evaluation is

documented in NUREG-1366, ``Improvements to Technical Specification

Surveillance Requirements,'' dated December 1992. The NRC staff found,

that while the majority of testing at power is important, safety can be

improved, equipment degradation decreased, and an unnecessary burden on

personnel resources eliminated by reducing the amount of testing at

power that is required by TS. Based on the results of the evaluations

documented in NUREG 1366, the NRC issued Generic Letter 93-05.

The Auxiliary Feedwater System supplies water to the steam

generators to remove decay heat from the Reactor Coolant System. To

ensure operability of the Auxiliary Feedwater System, the pumps are

currently tested on a monthly basis as required by the TS. Consistent

with Generic Letter 93-05, Item 9.1 and NUREG-1366, the licensee is

requesting a change to the surveillance testing frequency for the

Auxiliary Feedwater Pumps from monthly to quarterly on a staggered test

basis.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Specifically, operation of North Anna Power Station in

accordance with the proposed Technical Specifications changes will

not:

1. Involve a significant increase in the probability of

occurrence or consequences of an accident previously evaluated.

Changing the surveillance test frequencies of the Auxiliary

Feedwater System pumps does not significantly affect the probability

of occurrence or consequences of any previously evaluated accidents.

Quarterly testing of the pumps on a staggered basis will continue to

assure that the Auxiliary Feedwater System will be capable of

performing its intended functions. Therefore, the change in

frequency of testing the Auxiliary Feedwater System pumps does not

affect the probability or consequences of any previously analyzed

accident.

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

from any accident previously evaluated.

Changing the surveillance test frequency of the Auxiliary

Feedwater System pumps does not involve any physical modification of

the plant or result in a change in a method of operation. Quarterly

testing of the Auxiliary Feedwater System pumps on a staggered basis

will continue to assure that the Auxiliary Feedwater System will be

capable of performing its intended function. Therefore, a new or

different type of accident is not made possible.

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

Changing the surveillance test frequency of the Auxiliary

Feedwater System pumps does not affect any safety limits or limiting

safety system settings. System operating parameters are unaffected.

The availability of equipment required to mitigate or assess the

consequence of an accident is not reduced. Quarterly testing of the

Auxiliary Feedwater System pumps on a staggered basis will continue

to assure that the Auxiliary Feedwater System will be capable of

performing its intended functions. Safety margins are, therefore,

not decreased.

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

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.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

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

North Carolina

Date of application for amendment: October 5, 1993

Brief description of amendment: The proposed change to the

Technical Specifications would revise the wording of liquid release

rate limit and its associated bases, and relocate the old 10 CFR 20.106

requirements to the new 10 CFR 20.1302 to be consistent with the

revised terminology of 10 CFR Part 20. The new wording will retain the

same overall level of effluent control required to meet the design

objectives of Appendix I to 10 CFR Part 50.

Date of issuance: December 14, 1993

Effective date: December 14, 1993

Amendment No.: 40

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

Technical Specifications.

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

FR 59746) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 14, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: Cameron Village Regional

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

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

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

Date of application for amendments: August 5, 1992

Brief description of amendments: The amendment revises the

Braidwood Station, Units 1 and 2, Technical Specifications (TS)

regarding Engineered Safety Features Actuation System (ESFAS)

instrumentation. The ESFAS, Functional Units, Analog Channel

Operational Test interval is changed from monthly to quarterly.

Eighteen changes to the Reactor Trip System (RTS) are also included in

this TS change.

Date of issuance: December 16, 1993

Effective date: December 16, 1993

Amendment Nos.: 44 and 44

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 28, 1992 (57 FR

48816) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 16, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: Wilmington Township Public

Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

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

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

Date of application for amendments: October 29, 1993

Brief description of amendments: The amendments revise Table 3.6.3-

1, ``Primary Containment Isolation Valves,'' of the LaSalle Technical

Specifications for Units 1 and 2 by adding a new category of valves to

these tables. There are a total of eight new valves added in each

table, consisting of two check valves in each of four backfill lines.

The backfill lines were added in response to NRC Bulletin 93-03,

``Resolution of Issues Related to Reactor Vessel Water Level

Instrumentation in BWRs,'' dated May 28, 1993.

Date of issuance: December 10, 1993

Effective date: December 10, 1993

Amendment Nos.: 92 and 76

Facility Operating License Nos. NPF-11 and NPF-18. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1993 (58 FR

59493). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 10, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: Public Library of Illinois

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

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

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

Date of application for amendment: November 25, 1992, as

supplemented by letter dated February 5, 1993.

Brief description of amendment: The amendment revises surveillance

intervals for Process Radiation Monitors, Area Radiation Monitors, the

Main Steam Line Radiation Monitors, the Auxiliary Feedwater System

Initiating Logic, the Main Steam Safety Valves Setpoints, and the Toxic

Gas Detection System Monitors to accommodate a 24-month refueling

cycle. These revisions are being made in accordance with the guidance

provided by Generic Letter 91-04, ``Changes in Technical Specification

Surveillance Intervals to Accommodate a 24-Month Fuel Cycle.''

Date of issuance: December 16, 1993

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

30 days.

Amendment No.: 166

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 25, 1993 (58 FR

16219) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 16, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

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

Date of application for amendments: October 5, 1993, as

supplemented November 15 and 22, 1993

Brief description of amendments: The amendments revise the

Technical Specifications to reflect the appropriate operability

requirements for cold leg accumulator water volume and surveillance

requirements values for the centrifugal changing pumps, safety

injection pumps, and residual heat removal pumps to prevent possible

runout conditions during a loss of coolant accident event.

Date of issuance: December 15, 1993

Effective date: December 15, 1993

Amendment Nos.: 110 and 104

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1993 (58 FR

57848) The November 15 and 22, 1993, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 15, 1993. No significant hazards

consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of application for amendments: October 5 and 14, 1993, as

supplemented November 15 and December 14, 1993

Brief description of amendments: The amendments revise the

Technical Specifications to allow the implementation of interim steam

generator tube plugging criteria for the tube support plate elevations.

Date of issuance: December 16, 1993

Effective date: December 16, 1993

Amendment Nos.: 111 and 105

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1993 (58 FR

57849) The November 15 and December 14, 1993, letters provided

clarifying information and revisions to the coolant specific activity

that did not change the scope of the original application and did not

change the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 16, 1993. No significant hazards

consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of application for amendments: September 7, 1993

Brief description of amendments: The amendments revise the

Technical Specification to (a) reduce the slope of the axial power

imbalance penalty in the overtemperature-delta temperature reactor

protection system trip setpoint equation, and (b) increase the boron

concentration limits in the cold leg accumulators, the refueling water

storage tank, the reactor coolant system, and refueling canal during

MODE 6 conditions. These changes reflect the reloading of Unit 1 with

Mark BW fuel for Cycle 8 including an increase in cycle length from 350

effective full power days (EFPD) to 390 EFPD.

Date of issuance: December 17, 1993

Effective date: December 17, 1993

Amendment Nos.: 112 and 106

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1993 (58 FR

57847) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 17, 1993. No

significant hazards consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of application for amendments: October 25, 1993, as

supplemented December 3 and 6, 1993

Brief description of amendments: The amendments revise the

Technical Specifications (TS) Figure 2.1-1, certain TS Table 2.2-1

factors in the equation for the OVERTEMPERATURE delta T and OVERPOWER

delta T setpoints, and Figure 3.2-1 to reflect a reduction in the

required minimum measured reactor coolant system (RCS) flow rate from

385,000 gallons per minute (gpm) to 382,000 gpm for Unit 1. Catawba

Unit 2 values are unchanged and, accordingly, certain TS pages were

modified to retain the current TS values in effect for Unit 2.

The need for these changes is attributed to the effects of steam

generator tube plugging and to a hot leg temperature streaming

phenomenon. The application also proposed to revise the text of TS

2.1.1 and the definition for TS Figure 2.1-1. These changes are not

related to the changes in RCS flow rate. The staff is continuing to

review these proposed changes and, accordingly, they are not dealt with

in this amendment.

Date of issuance: December 17, 1993

Effective date: Effective within 30 days of its date of issuance.

Amendment Nos.: 113 and 107

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

revised the Technical Specifications.

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

FR 59747) The December 3 and 6, 1993, letters provided clarifying

information that did not change the scope of the October 25, 1993,

application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 17, 1993. No significant hazards

consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Mississippi Power &

Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit

1, Claiborne County, Mississippi

Date of application for amendment: May 20, 1993

Brief description of amendment: The amendment removed unnecessary

operability requirements for the Intermediate Range Monitors (IRMs) and

the Average Power Range Monitors (APRMs) during plant shutdown

operations.

Date of issuance: December 13, 1993

Effective date: December 13, 1993

Amendment No: 109

Facility Operating License No. NPF-29. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 23, 1993 (58 FR

34077) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 13, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: Judge George W. Armstrong

Library, Post Office Box 1406, S. Commerce at Washington, Natchez,

Mississippi 39120.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of application for amendments: July 23, 1993

Brief description of amendments: The amendments are necessary to

implement new Standards for Protection Against Radiation (10 CFR Part

20).

Date of issuance: December 16, 1993

Effective date: December 16, 1993

Amendment Nos.: 125, 63

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the Technical Specifications.

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

FR 46234) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 16, 1993. No

significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: October 8, 1993

Brief description of amendment: The amendment deletes portions of

the Oyster Creek Nuclear Generating Station Radiological Effluent

Technical Specifications and relocates them to controlled programs in

accordance with the guidance contained in NRC Generic Letter 89-01,

dated January 31, 1989.

Date of issuance: December 13, 1993

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

30 days.

Amendment No.: 166

Facility Operating License No. DPR-16. Amendment revised the

Technical Specifications.

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

FR 59749) The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated December 13, 1993. No

significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, New Jersey

08753.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: October 18, 1993

Brief description of amendment: The amendment revises the Technical

Specifications to delete requirements to demonstrate by testing, that a

redundant system/component is operable when a system/component is

declared inoperable. In lieu of testing the redundant system/component

to demonstrate its operability the Technical Specifications are being

revised to require an administrative check of plant records to verify

operability of the redundant system/component. Confirming changes are

made to Definition 1.1 ``Operable-Operability.''

Date of issuance: December 21, 1993

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

60 days.

Amendment No.: 167

Facility Operating License No. DPR-16. Amendment revised the

Technical Specifications.

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

FR 59749) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 21, 1993. No

significant hazards consideration comments received: Yes. Comments were

provided by letter dated December 10, 1993, from the State of New

Jersey, Department of Environmental Protection and Energy, Division of

Environmental Safety, Health and Analytical Programs. The comments and

the NRC staff's response are addressed in the Commission's Safety

Evaluation dated December 21, 1993.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, New Jersey

08753

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile

Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: September 20, 1993, as

supplemented on October 1, 1993.

Brief description of amendment: The amendment revises the plant

Technical Specifications to reflect a partial GPU Nuclear

reorganization to become effective when Three Mile Island, Unit 2 (TMI-

2), enters the Post-Defueling Monitored Storage (PDMS) mode. This

reorganization includes deleting TMI-2 as a Division and incorporating

those functions and responsibilities required to maintain the PDMS

condition and requirements into the current TMI-1 Division. The TMI-1

Division will be renamed the TMI Division.

Date of issuance: December 13, 1993

Effective date: No specific date has been specified by the staff

for the effectiveness of this amendment. The amendment will become

fully effective at such time as the Vice President - TMI has been

delegated the full responsibility of the overall safe operation of both

TMI-1 and TMI-2.

Amendment No.: 179

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 13, 1993 (58 FR

52987). The October 1, 1993, submittal provided clarifying and

corrected TS pages which did not change the initial proposed no

significant hazards consideration determination. The Commission's

related evaluation of this amendment is contained in a Safety

Evaluation dated December 13, 1993. No significant hazards

consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile

Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: August 9, 1993

Brief description of amendment: The amendment revises the plant

Technical Specifications to be consistent with a major revision to 10

CFR Part 20 that is to be implemented by January 1, 1994.

Date of issuance: December 21, 1993

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

January 1, 1994.

Amendment No.: 180

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

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

FR 59751). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 21, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile

Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: August 24, 1993.

Brief description of amendment: The amendment revises the plant

Technical Specifications to adopt the Standard Technical Specification

(STS) provision that allows a period up to 24 hours to complete a

surveillance requirement upon the discovery that the surveillance has

been missed.

Date of issuance: December 22, 1993

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

within 30 days of issuance.

Amendment No.: 181

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

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

FR 59751). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 22, 1993. No

significant hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

Gulf States Utilities Company and Cajun Electric Power Cooperative,

Docket No. 50-458, River Bend Station, Unit 1, West Feliciana

Parish, Louisiana

Date of amendment request: January 13, 1993, as supplemented by

letter dated October 18, 1993.

Brief description of amendment: The amendment revises the River

Bend, Unit 1 operating license to reflect a change in ownership of Gulf

States Utilities (GSU). GSU, which owns a 70 percent undivided interest

in the River Bend Station, will become a wholly-owned subsidiary

company of Entergy Corporation.

Date of issuance: December 16, 1993

Effective date: December 6, 1993, to be implemented within 180 days

of issuance.

Amendment No.: Amendment No. 69

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

license.

Date of initial notice in Federal Register: July 7, 1993 (58 FR

36435) The October 18, 1993, supplemental letter provided additional

clarifying information and did not change the initial no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

December 16, 1993. No significant hazards consideration comments

received: Yes. Comments and a request for hearing were received from

Cajun Electric Power Cooperative of Baton Rouge, Louisiana.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

Gulf States Utilities Company, Cajun Electric Power Cooperative,

and Entergy Operations, Inc., Docket No. 50-458, River Bend

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: January 13, 1993, as supplemented by

letter dated June 29, 1993.

Brief description of amendment: The amendment revises the River

Bend Station, Unit 1 operating license to include as a licensee,

Entergy Operations, Inc. (EOI), and to authorize EOI to use and operate

River Bend and to possess and use related licensed nuclear materials.

Date of issuance: December 16, 1993

Effective date: December 16, 1993 to be implemented within 180 days

of issuance.

Amendment No.: 70

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

license.

Date of initial notice in Federal Register: July 7, 1993 (58 FR

36436) The June 29, 1993, supplemental letter provided additional

clarifying information and did not change the initial no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

December 16, 1993. No significant hazards consideration comments

received: Yes. Comments and a request for hearing were received from

Cajun Electric Power Cooperative of Baton Rouge, Louisiana.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

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: June 18, 1993

Brief description of amendment: The proposed changes to Technical

Specifications 6.2.3.1, ``Independent Safety Engineering Group (ISEG)

Function;'' 6.2.3.4, ``ISEG Records;'' 6.4.1, ``Training;'' and

6.5.2.2, ``Nuclear Review and Audit Group (NRAG) Composition'' are

editorial changes reflecting recent administrative/organizational

changes which occurred at Clinton Power Station.

Date of issuance: November 29, 1993

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

Amendment No.: 86

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

Technical Specifications.

Date of initial notice in Federal Register: August 18, 1993 (58 FR

43927) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 29, 1993. 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

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

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

Date of application for amendment: April 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.

Date of issuance: December 22, 1993

Effective date: December 22, 1993

Amendment No.: 158

Facility Operating License No. DPR-74. Amendments revised the

Technical Specifications.

Date of initial notice in Federal Register: August 4, 1993 (58 FR

41505) The supplemental letters provided clarifying information which

did not change the staff's initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated December 22, 1993.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile

Point Nuclear Station, Unit 2, Oswego County, New York

Date of application for amendment: June 7, 1993

Brief description of amendment: The amendment revises Technical

Specification (TS) 3/4.8.1, ``AC Sources-Operating,'' and associated

Bases to eliminate unnecessary diesel generator testing when a diesel

generator or an offsite power source becomes inoperable. The amendment

is intended to increase diesel generator reliability and the overall

level of plant safety by reducing the stresses on the diesel generators

caused by unnecessary testing. The amendment also makes additional

changes to TS 3/4.8.1 to further enhance diesel generator reliability

and incorporate certain administrative changes.

Date of issuance: December 15, 1993

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

30 days.

Amendment No.: 54

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

Technical Specifications.

Date of initial notice in Federal Register: July 7, 1993 (58 FR

36440) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 15, 1993. No

significant hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

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

York 13126.

North Atlantic Energy Service Corporation, Docket No. 50-443,

Seabrook Station, Unit No.

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