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 April 13, 1996, through April 26, 1996. The

last biweekly notice was published on April 24, 1996 (61 FR 18162).

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 20555-0001, and should cite the publication date and

page number of this Federal Register notice. Written comments may also

be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.

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

intervene is discussed below.

By June 7, 1996, 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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, Attention: Docketing and

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

Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,

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 20555-0001,

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,

Maryland

Date of amendments request: March 28, 1996.

Description of amendments request: Pursuant to 10 CFR 50.90, the

Baltimore Gas and Electric Company (BGE) hereby requests an amendment

to Operating License Nos. DPR-53 and DPR-69 to reduce the moderator

temperature coefficient (MTC) limit shown on Technical Specification

Figure 3.1.1-1. This proposed change is necessary to support changes in

the safety analyses made to accommodate a larger number of plugged

steam generator (SG) tubes for future operating cycles. The proposed

limit will be more restrictive than the existing limit to match the

analytical assumptions. In addition, the licensee provided information

to clarify the relationship of the MTC to an Anticipated Transient

Without Scram event in its licensing 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:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The safety analyses for the current fuel cycles assume 500 tubes

per steam generator (SG) are plugged and the maximum beginning-of-

cycle moderator temperature coefficient (MTC) is assumed to follow

the curve in Technical Specification Figure 3.1.1.-1. For the fuel

cycle to be installed in Unit 1 in spring 1996, Baltimore Gas and

Electric Company (BGE) assumes in the analyses that more SG tubes

are plugged than the current limit, and it is necessary to credit a

more restrictive (less positive) limit on the maximum positive MTC

to mitigate the

[[Page 20844]]

Reactor Coolant System pressure and temperature increase analyzed

for these events. Therefore, we are proposing a change to the

allowable positive MTC limits shown on Technical Specification

Figure 3.1.1-1. The proposed limit will be more restrictive than the

existing limit to match the analytical assumptions. Since the safety

analyses supporting an increase in the number of plugged SG tubes

are applicable to both Units 1 and 2, BGE is requesting this change

for both Units.

The proposed change makes the limit on the maximum positive MTC

more restrictive. From an operational standpoint, a more restrictive

limit on MTC will help mitigate the effect of plant transients on

control of plant parameters (e.g., reactor power, pressurizer

pressure, pressurizer level, etc.) Therefore, the probability of a

previously analyzed accident will not be significantly increased.

The reason for the proposed change is to mitigate the effect

(increased reactor coolant temperatures) of increased SG U-tube

plugging on the results of the affected safety analyses. Using the

more restrictive limit on the maximum positive MTC, the Loss of

Load, Loss of Feedwater Flow, Feed Line Break, and Control Element

Assembly Withdrawal events were reanalyzed using previously accepted

methodologies. The results of these analyses are within the

acceptance limits for these events. Therefore, the consequences of a

previously analyzed accident will not be significantly increased.

The proposed change is similar to the examples of amendments

that are considered not likely to involve significant hazards

considerations given in the Statements of Consideration for 10 CFR

50.92 (51 FR 7744). The example of interest is, ``A change that

constitutes an additional limitation, restriction, or control not

presently included in the technical specifications, e.g., a more

stringent surveillance requirement.'' The proposed change provides a

more restrictive limit on the positive MTC given in Technical

Specification Figure 3.1.1-1. Based on the above arguments and the

similarity to an example in the Federal Register, BGE has determined

that the proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

The proposed change makes the limit on the maximum positive MTC

more restrictive. The proposed change does not involve installation

of new or different equipment, modify the interfaces with existing

equipment, change the equipment's function, or change the method of

operating the equipment. The proposed change does not affect normal

plant operations or configurations. The more restrictive MTC limit

will help mitigate the effect of plant transients on control of

plant parameters.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

safety.

The proposed change provides for a more restrictive limit for

the allowable positive MTC. The more restrictive limit on the

maximum positive MTC was evaluated using previously approved

methodologies and compared to the existing acceptance criteria. The

analyses show that the proposed change preserves the margin of

safety by ensuring that the results of the safety analyses for the

Loss of Load, Loss of Feedwater Flow, Feed Line Break, and Control

Element Assembly Withdrawal events meet established NRC acceptance

limits for these events.

In addition, this proposed change is similar to the example of

amendments that are considered not likely to involve significant

hazards considerations given in the Statements of Consideration for

10 CFR 50.92 (51 FR 7744). The example of interest is, ``A change

that constitutes an additional limitation, restriction, or control

not presently included in the technical specifications, e.g., a more

stringent surveillance requirement.'' The proposed change provides a

more restrictive limit on the positive MTC given in Technical

Specification Figure 3.1.1-1. Based on the above arguments and the

similarity to an example in the Federal Register, BGE has determined

that the proposed change does not involve a significant reduction in

the margin of safety.

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

this review, it appears that the three standards of 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: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: Susan F. Shankman, Acting.

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

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

Carolina

Date of amendment request: February 8, 1996.

Description of amendment request: The proposed amendment would

remove Technical Specifications (TS) 3.3.4, Turbine Overspeed

Protection; TS 3.7.12, Area Temperature Monitoring; and TS 3.11.2.6,

Gas Storage Tanks; and their associated bases; and relocate them to

licensee-controlled documents, such as the Final Safety Analysis

Report. The licensee revised the original amendment request dated

October 24, 1994, to provide supplemental information to TS 6.8.4 for

administrative control program related to TS 3.11.2.6, by letters dated

August 31, 1995 and February 8, 1996.

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 was previously presented in the Federal Register

(59 FR 60397). The staff reviewed and determined that the proposed

license amendment's revisions do not alter the original conclusion that

no significant hazards considerations exist pursuant to 10 CFR 50.92.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Eugene V. Imbro.

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

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

Date of amendment request: December 21, 1995.

Description of amendment request: The proposed amendments would

delete the requirement to place the reactor mode switch in the Shutdown

position if a stuck open safety/relief valve cannot be closed within

two minutes. The operator would still be required to scram the reactor

if suppression pool average water temperature reaches 110 degrees

Fahrenheit or greater. The licensee also proposed changes to the TS

index pages to reflect Bases page changes that were accepted by the NRC

staff in a letter dated May 23, 1995. Because the changes to the index

pages require a license amendment, they have been included as part of

this submittal.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The proposed change does not involve a significant increase in

the probability or

[[Page 20845]]

consequences of an accident previously evaluated in the UFSAR. A

stuck open SRV event is a mild transient which neither affects fuel

limits nor radiological consequences. The two minute requirement to

manually scram after a SRV becomes stuck open is not assumed or used

in any transient or accident analysis in the FSAR. Removing the two

minute requirement to manually scram after a SRV becomes stuck open

does not change the probability of any accident evaluated in the

FSAR. Removing the two minute requirement to manually scram after a

SRV becomes stuck open also does not change the capability of the

suppression pool during this event in case of any accident involving

reactor blowdown, because the suppression pool average water

temperature limit in Technical Specification 3.6.2.1 is still valid

and enforced. The suppression pool average water temperature limit

is the only requirement during operational conditions 1 and 2 that

assures sufficient heat sink capacity in case of a LOCA in the

containment. Therefore, removing the two minute requirement to

manually scram after a SRV becomes stuck open would not increase the

probability or consequences of any postulated accident analyzed in

the FSAR.

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

accident from any accident previously evaluated because:

The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated

in the UFSAR. This change does not effect any hardware. This is a

procedural change to assure that the reactor will not be

unnecessarily scrammed by the operator after a SRV is stuck open for

two minutes. The reactor will still be scrammed if suppression pool

average water temperature increases above 110 degrees F. Since the

design basis of the suppression pool is protected by this average

water temperature limit, this procedural change of removing the two

minute requirement to manually scram after a SRV becomes stuck open

introduces no new accident or malfunction.

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

because:

The proposed change does not reduce the margin as defined in the

bases for any Technical Specification. On the contrary, if the two

minute requirement to manually scram after a SRV becomes stuck open

is not removed, the operator has to scram the reactor thus

challenging the RPS, the rector vessel, and other associated

components, and reducing the related margin to safety. This scram

would be unnecessary if the suppression pool average water

temperature is below the 110 degree F limit allowed by the design

basis of the suppression pool. Reactor safety or suppression pool

design basis is not compromised because the suppression pool average

water temperature limit alone guarantees that there would not be any

reduction in margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

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

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

Date of amendment request: March 4, 1996.

Description of amendment request: The proposed amendments would

change the McGuire Units 1 and 2 Updated Final Safety Analysis Report

to delete the seismic qualification requirement for the Containment

Atmosphere Particulate Radiation Monitors.

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:

This proposed change has been evaluated against the standards in

10 CFR 50.92 and has been determined to involve no significant

hazards considerations, in that operation of the facility in

accordance with the proposed amendment would not:

1. [I]nvolve a significant increase in the probability or

consequences of an accident previously evaluated; or

EMF38(L) is not used directly for any phase of power generation

or conversion or transmission, normal decay heat removal, fuel

handling, or the processing of radioactive fluids. As such, it is

not an ``accident initiator''. No ``accident initiator'' is affected

by the change. Thus, the probability of accidents evaluated in the

FSAR is not affected by the change. It is determined that sufficient

ability to determine conditions inside containment remain available

for any earthquake up to and including the SSE [safe-shutdown

earthquake]. Furthermore, should either EMF38(L) or EMF39(L) be

found to not be functional following any earthquake, including those

smaller than the OBE [Operating Basis Earthquake], the appropriate

steps will be taken; i.e., declare the monitor(s) inoperable and

apply the action statement for TS [technical specification] 3.4.6.1

which may require that the associated unit(s) be taken to Cold

Shutdown (Mode 5) if the minimum required Reactor Coolant Leakage

Detection Systems are not operable. Cold Shutdown is a mode for

which neither the Emergency Core Cooling System nor the containment

safeguards are required. Finally, no equipment provided to mitigate

any accident is adversely affected by the change. For these reasons,

the proposed change will not involve a significant increase in the

probability or consequences of an accident previously evaluated in

the SAR [safety analysis report].

2. [C]reate the possibility of a new or different type of

accident from any accident previously evaluated; or

As stated above, no equipment used in direct support of power

generation or conversion or transmission, normal decay heat removal,

fuel handling, or processing of radioactive fluids is affected with

the update. No new failure modes are identified with the change. The

upper bound to an undetected leak in the Reactor Coolant System is a

Loss of Coolant Accident [LOCA]. As noted above, no equipment

provided to mitigate a LOCA is affected by the change. For these

reasons, the change will not create a new or different type of

accident from any accident previously evaluated.

3. [I]nvolve a significant reduction in a margin of safety.

It has been determined that sufficient means remain at the

disposal to the operators to assess conditions within the

containment following any earthquake up to and including the SSE. In

particular, the ability to determine leakage with the sensitivity

comparable to that of EMF38(L) can be established. This meets the

intent of the Regulatory Position of RG [Regulatory Guide] 1.45. In

addition, should it be determined that either EMF38(L) or EMF39(L)

is not functional following any earthquake, the appropriate steps

will be taken; i.e, declare the monitor(s) inoperable and apply the

action statement for TS 3.4.6.1 which may require that the

associated unit(s) be taken to Cold Shutdown (Mode 5) if the minimum

required Reactor Coolant Leakage Detection Systems are not operable.

This brings the unit(s) to a mode in which TS 3.4.6.1 does not

apply. It ensures that at least the minimum required Reactor Coolant

System leakage detection systems will be functional before power

operations are continued following a postulated earthquake smaller

than the OBE. It ensures protection of the reactor coolant pressure

boundary, one of the fission product barriers. No other fission

product barrier is affected by the change. Therefore, the margin of

safety is not reduced.

Therefore, based on the information contained in this submittal,

it is determined that no significant hazard is associated with the

proposed change.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

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

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

South

[[Page 20846]]

Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas

Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas

Date of amendment request: April 11, 1996.

Description of amendment request: The proposed technical

specification amendment modifies the reactor building leak testing

requirements per Option B to 10 CFR 50, Appendix J. Option B permits

performance based determination of the reactor building leak testing

frequency.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does Not Involve a Significant Increase in the Probability or

Consequences of an Accident Previously Evaluated.

The proposed changes to the Technical Specifications implement

Option B of 10 CFR 50 Appendix J at ANO. The proposed changes will

result in increased intervals between containment leakage tests

determined through a performance based approach. The intervals

between such tests are not related to conditions which cause

accidents. The proposed changes do not involve a change to the plant

design or operation. Therefore, this change does not involve a

significant increase in the probability of any accident previously

evaluated.

NUREG-1493, ``Performance-Based Containment Leak-Test Program,''

contributed to the technical bases for Option B of 10 CFR 50

Appendix J. NUREG-1493 contains a detailed evaluation of the

expected leakage from containment and the associated consequences.

The increased risk due to lengthening of the intervals between

containment leakage tests was also evaluated and found acceptable.

Using a statistical approach, NUREG-1493 determined the increase in

the expected dose to the public from extending the testing frequency

is extremely small. It also concluded that a small increase is

justifiable due to the benefits which accrue from the interval

extension. The primary benefit is in the reduction in occupational

exposure. The reduction in the occupational exposure is a real

reduction, while the small increase to the public is statistically

derived using conservative assumptions. Therefore, this change does

not involve a significant increase in the consequences of any

accident previously evaluated.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

2. Does Not Create the Possibility of a New or Different Kind of

Accident from any Previously Evaluated.

The proposed change to the Technical Specifications incorporates

the performance based approach authorized by Option B of 10 CFR 50

Appendix J. The interval extensions allowed by this change do not

involve a change to the plant design or operation. No safety related

equipment or safety functions are altered as a result of this

change. The reduced testing frequency does not affect the testing

methodology. As a result, the proposed change does not affect any of

the parameters or conditions that could contribute to initiation of

any accidents. No new accident modes are created by extending the

test intervals. Therefore, this change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. Does Not Involve a Significant Reduction in the Margin of

Safety.

The proposed change does not change the performance methodology

of the containment leakage rate testing program. However, the

proposed change does affect the frequency of containment leakage

rate testing. With an increased frequency between tests, the

proposed change does increase the probability that a increase in

leakage could go undetected for a longer period of time. Operational

experience has demonstrated the leak tightness of the containment

buildings has been significantly below the allowable leakage limit.

The margin to safety that has the potential of being impacted by

the proposed change involves the offsite dose consequences of

postulated accidents which are directly related to containment

leakage rates. The limitation on containment leakage rate is

designed to ensure the BWN total leakage volume will not exceed the

value assumed in our accident analysis. The margin to [sic] safety

for the offsite dose consequences of postulated accidents directly

related to containment leakage is maintained by meeting the 1.0 L.

acceptance criteria. The proposed change maintains the 1.0 L.

acceptance criteria.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.

NRC Project Director: William D. Beckner.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: April 11, 1996.

Description of amendment request: The proposed technical

specification (TS) amendment adds low-temperature overpressure

protection (LTOP) requirements to the TSs to resolve Generic Issue 94

in accordance with Generic Letter 90-06.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does Not Involve a Significant Increase in the Probability or

Consequences of an Accident Previously Evaluated.

This proposed change provides additional controls in the ANO-2

Technical Specification [(TS)] for ensuring that LTOP [low-

temperature overpressure protection] protection is available when

required. The limiting condition involving the simultaneous

injection of two HPSI [high pressure safety injection] and three

charging pumps to an RCS [reactor coolant system] water solid

condition, was used in the calculation of the ANO-2 proposed LTOP

setpoints. The methodology utilized in the LTOP setpoint analysis is

based on ASME [American Society of Mechanical Engineers] Code Case

N-514. The code case establishes a factor of 110 percent of the

operating pressure temperature curves instead of 100 percent. The

safety factor utilized by the code case provides a more reasonable

vessel overpressure allowance for conditions expected under pressure

loading from low temperature transients. The SITs [safety injection

tanks] are required to be isolated, if not depressurized, prior to

entering the LTOP enable temperature and are periodically verified

to be isolated when LTOP conditions exist. The LTOP setpoint of the

relief valves proposed by this technical specification [TS] change

is not considered to be an initiator of any transients, but is used

to mitigate an overpressure condition if such a transient were to

occur.

Therefore, this change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

2. Does Not Create the Possibility of a New or Different Kind of

Accident from any Previously Evaluated.

The design basis event for establishing LTOP limits is the

simultaneous injection of two HPSI and three charging pumps to an

RCS water solid condition. The LTOP vent size of 6.38 square inches

and the valve pressure setpoint of less than or equal to 430 psig

are currently used for mitigation of low temperature overpressure

conditions. The change in the enable setpoint was analyzed by the

application of Code Case N-514 and determined to adequately ensure

that this temperature [sic] setpoint will mitigate a

[[Page 20847]]

LTOP transient. The operator action to enable the LTOP relief valves

at 220 degrees ensures that the RCS including the reactor vessel

will not undergo system pressures at low temperature conditions

beyond their design limits. Therefore, there will not be any impact

to systems, structures or components beyond their design

requirements.

Therefore, this change does not create the possibility of a new

or different kind of accident from any previously evaluated.

3. Does Not Involve a Significant Reduction in the Margin of

Safety.

The addition of a new specification to the ANO-2 Technical

Specification [TS] will not significantly reduce the margin of

safety. The LTOP safety factors are based on reanalyzed conditions

for 21 effective full power years (EFPY) of operation utilizing

methodology contained in ASME Code Case N-514. The LTOP evaluation

under Code Case N-514 for low temperature transients is considered

more appropriate than the ASME Section XI. The code case establishes

a factor of 110 percent of the operating pressure temperature curves

instead of 100 percent. The safety factor utilized by the code case

provides a more reasonable vessel overpressure allowance for

conditions expected under pressure loading from low temperature

transients. Although the proposed setpoint may involve a slight

reduction in a margin of safety, the enable temperature setpoint

will provide an equivalent level of safety to the reactor vessel

during LTOP transients and will satisfy the purpose of 10 CFR 50.60

for fracture toughness. Therefore, based on the refined methodology

used to calculate ANO-2 LTOP setpoints for 21 EFPY the margin of

safety will not be significantly reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.

NRC Project Director: William D. Beckner.

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: April 18, 1996.

Description of amendment request: The licensee has proposed to

delete a restriction on the 24-hour emergency diesel generator

operation test in Surveillance Requirement 3.8.1.14 (Page 3.8-12) of

the Technical Specifications (TSs) for the Grand Gulf Nuclear Station,

Unit 1. The deletion would allow the test to also be conducted during

power operation (i.e., during Modes 1 and 2), instead of the current

requirement to only conduct the test when the plant is shut down.

The frequency of conducting this test, the conditions of the test,

and the criteria to pass the test are not being changed by this

amendment request.

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 for the amendment request, which is presented below:

Entergy Operations, Inc. [(EOI)] propose[d] to change the

current Grand Gulf Nuclear Station [GGNS] Technical Specifications

[(TSs)]. The specific change is to modify note 2 to Surveillance

3.8.1.14. Presently, this note prohibits the performance of the 24

hour diesel maintenance run while the unit is in either Mode 1 or 2.

The proposed change would remove this restriction thus allowing the

24 hour run to be performed during any mode of operation (i.e.,

modes 1, 2, 3, 4 or 5).

The Commission has provided standards for determining whether a

no significant hazards considerations exists as stated in 10 CFR

50.92 (c). A proposed amendment to an operating license involves no

significant hazards consideration if operation of the facility in

accordance with the proposed amendment would not: (1) involve a

significant increase in the probability or consequences of an

accident previously evaluated; (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.

Entergy Operations, Inc. [EOI] has evaluated the no significant

hazards consideration in its request for this license amendment and

determined that no significant hazards considerations results from

this change. In accordance with 10 CFR 50.91(a), Entergy Operations,

Inc. [EOI] is providing the analysis of the proposed amendment

against the three standards in 10 CFR 50.92(c). A description of the

no significant hazards consideration determination follows:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

The GGNS UFSAR [Updated Final Safety Analysis Report] assumes

that the AC electrical power sources are designed to provide

sufficient capacity, capability, redundancy and reliability to

ensure that the fuel, reactor coolant system and containment design

limits are not exceeded during an assumed design basis event.

Specifically, the UFSAR assumes that the onsite EDG's [emergency

diesel generator's] provide emergency power in the event offsite

power is lost to either one or all three ESF [engineered safety

feature] buses. In the event of a loss of preferred power, the ESF

electrical loads are automatically connected to the EDG's in

sufficient time to provide for safe reactor shutdown and to mitigate

the consequences of a design basis accident such as a LOCA.

The proposed change to permit the 24 hour testing of the EDG's

during power operation does not increase the chances or consequences

of any previously evaluated accident. The capability of the EDG's to

supply power in a timely manner will not be compromised by

permitting performance of EDG testing during periods of power

operation. Design features of the EDG's and electrical systems

ensures that if a LOCA [loss of coolant accident] or LOP [loss of

offsite power] signal, either individually or concurrently, should

occur during testing that the EDG would be returned to its ready-to-

load operation (i.e., EDG running at rated speed and voltage

separated from the offsite sources) or separately connected to the

ESF bus providing ESF loads. As such, an EDG being tested is

considered to be Operable and fully capable of meeting its intended

design function. Additionally, the testing of an EDG is not a

precursor to any previously evaluated accidents.

Therefore, the proposed change allowing testing of EDG's during

power operation will not significantly increase the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

As previously discussed [above], the proposed change to permit

the performance of EDG testing during power operation will not

affect the operation of any system or alter any system's response to

previously evaluated design basis events. The EDG's will

automatically transfer from the test configuration to the ready-to-

load configuration following receipt of a valid signal (i.e., LOCA

or LOP). In the ready-to-load configuration, the EDG will be running

at rated speed and voltage separated from the offsite source capable

of automatically supplying power to the ESF buses in the event that

preferred power is actually loss.

Surveillance Requirement 3.8.1.17 demonstrates that the EDG will

automatically override the test mode following generation of a LOCA

signal. In addition the ability of the EDG's to survive a full load

reject is verified by the performance of surveillance requirement

3.8.1.10. These existing surveillance requirements along with system

design features ensures that the performance of EDG testing during

power operation will not create the possibility of a new or

different kind of accident from any previously evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

The AC electrical power sources are designed to provide

sufficient capacity, capability, redundancy, and reliability to

ensure the availability of necessary power to ESF systems so that

the fuel, reactor coolant system and containment design limits are

not exceeded. Specifically, the EDG's must be capable of

automatically providing power to ESF loads in sufficient time to

provide for safe reactor shutdown and to mitigate the consequences

of a design basis accident in the event of a loss of preferred

power.

[[Page 20848]]

Testing of EDG's during power operation will not affect the

availability or operation of any offsite source of power. In

addition, the EDG being tested remains capable of meeting its

intended design functions. Therefore the proposed change to the

Technical Specification Surveillance Requirement 3.8.1.14 will not

result in 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: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120.

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502.

NRC Project Director: William D. Beckner.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: April 15, 1996 (TSCR No. 244).

Description of amendment request: The proposed amendment would

revise Specification 5.3.1.B of the Oyster Creek Technical

Specifications. The current specification prohibits handling a load

greater in weight than one fuel assembly over irradiated fuel in the

spent fuel storage facility. The proposed change will facilitate the

off load of spent fuel to the Oyster Creek Independent Spent Fuel

Storage Installation (ISFSI). Specifically, the shield plug for the dry

shield canister (DSC) and the associated lifting hardware will be moved

over irradiated fuel which is contained in the DSC within the transfer

cask located in the Cask Drop Protection System (CDPS).

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. State the basis for the determination that the proposed

activity will or will not increase the probability of occurrence or

consequences of an accident.

The design features and capacity of the reactor building crane

provide a significant safety factor. In addition, personnel training

and other administrative controls further reduce risk. Thus, the

dropping of the DSC shield plug onto a loaded DSC and causing damage

to the spent fuel assemblies is not a credible event. Therefore, it

does not increase the probability of or consequences of an accident.

2. State the basis for the determination that the activity does

or does not create the possibility of an accident or malfunction of

a different type than any previously identified in the SAR [safety

analysis report].

This activity will not create the possibility of a new or

different type of accident than previously evaluated in the SAR

because the proposed heavy load handling exception does not create a

new credible accident scenario. Dropping the shield plug on a loaded

DSC and damaging spent fuel assemblies is not considered a credible

event.

3. State the basis for the determination that the margin of

safety is not reduced.

This activity will not involve a significant reduction in the

margin of safety because the proposed heavy load handling evolution

does not create a credible accident scenario.

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: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

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.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of amendment request: April 19, 1996.

Description of amendment request: The proposed amendment would

revise Technical Specification 5.14 to add the appropriate references

identifying the detailed methodology and conditions for analyzing the

Small Break Loss-of-Coolant Accident (SBLOCA) to the list of the

approved Core Operating Limits Report methods.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the Proposed Amendment involve a significant increase in

the probability or consequences of an accident previously evaluated?

These Proposed Changes are administrative in nature and are

consistent with the guidance set forth in the NRC Generic Letter 88-

16 identifying the requirements for the inclusion of analytical

methodology references in Technical Specifications as used in

determining compliance with the regulatory limits.

The references, as proposed to be included in section 5.14 of

the Technical Specifications, have previously been reviewed and

approved by the NRC for generic applicability to PWRs [Pressurized

Water Reactors]. The reports identified in the Proposed Change have

been accepted by the NRC for referencing in plant licensing

applications.

Since the references listed in the Proposed Change have

previously been found to meet the conditions of 10 CFR 50.46 and 10

CFR Appendix K, and that the plant specific safety analysis

acceptance limits have not changed or been modified, the use of

these references in the analysis of SBLOCA accident for the Maine

Yankee plant is consistent with prior plant specific and industry

requirements and practices.

Therefore, we have concluded that the Proposed Change will not

result in a significant increase in the probability or consequences

of an accident previously evaluated.

2. Does the Proposed Amendment create the possibility for a new

or different kind of accident?

The Proposed Changes introduce no new mode of plant operation;

do not involve the physical modification of any structure, system,

or component; do not affect the function, operation or surveillance

for any equipment necessary for safe operation or shutdown of the

plant; and, do not involve any changes to setpoints or limits or

operating parameters. The Proposed Changes are administrative in

nature only.

Therefore, we have concluded that the Proposed Change cannot

result in the possibility of a new or different kind of accident

from that previously evaluated.

3. Does the Proposed Amendment involve a significant reduction

in a margin of safety?

The Proposed Changes are administrative in nature, consistent

with the guidance of Generic Letter 88-12, and have been reviewed

previously by the NRC and found acceptable with regard to the

requirements of 10 CFR 50.46 and 10 CFR Appendix K. Additionally,

the plant specific safety analysis acceptance criteria has not

changed from that used in the latest core reload analysis.

Therefore, we have concluded that 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: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME 04578.

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road, Brunswick, ME 04011.

[[Page 20849]]

NRC Deputy Director: John A. Zwolinski.

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

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

Date of amendment request: February 7, 1996.

Description of amendment request: The amendment would change the

operating license, the Technical Specifications, and associated Bases

to permit the use of 10 CFR Part 50, Appendix J, Option B, Performance-

Based Containment Leakage Rate Testing in accordance with the

implementation guidance in NRC's Regulatory Guide 1.163 dated September

1995. The change to the operating license would delete, in paragraph

2.D.ii, reference to certain exemptions to Appendix J previously

granted by the NRC, which would no longer be applicable once Option B

is implemented.

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 operation of Nine Mile Point Unit 2, in accordance with

the proposed amendment, will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

NMP2 [Nine Mile Point, Unit 2] is currently implementing Option

A of Appendix J of 10 CFR 50 for Type A, B and C testing. The

proposed change to the Operating License, the Technical

Specifications and the Bases would implement Option B to Appendix J

of 10 CFR 50 at NMP2 for Type A, B and C testing. Option B would

allow increased testing intervals after satisfying certain

performance based criteria. The proposed change also corrects an

inconsistency between the restoration statements and the

applicability requirements of LCO [Limiting Condition of Operation]

3.6.1.2. In addition, the proposed change affects the testing

intervals for the verification of the interlocks on the primary

containment air lock and for the measuring of the Hydrogen

Recombiner System leakage rate.

Appendix J describes the requirements for leakage testing of the

primary containment and its components penetrating the primary

containment. The leakage testing interval of the primary containment

and its components is not a precursor or initiator to an accident.

The primary containment and its penetrations minimizes the leakage

of radioactivity into the environment during an accident which

pressurizes the primary containment.

The testing intervals of the air lock interlocks and of the

Hydrogen Recombiner System leakage rate are also not precursors or

initiators to an accident. The interlocks function to provide

assurance that at least one air lock door will be closed and thereby

perform its accident mitigating function of minimizing the leakage

of radioactivity into the environment during accident conditions.

The Hydrogen Recombiner System is manually initiated following a

loss-of-coolant accident (LOCA) to maintain the hydrogen

concentration within the primary containment below its flammable

limit during post-LOCA conditions.

An inconsistency exists between the applicability statement of

LCO 3.6.1.2 and the requirement of the restoration statements to

restore prior to increasing reactor coolant system temperature over

200 deg.F. Eliminating this inconsistency does not diminish the

requirements contained in the Technical Specifications.

Therefore, the proposed change does not involve a significant

increase in the probability of an accident previously evaluated.

The proposed change to the Operating License, the Technical

Specifications and the Bases would replace the detailed and

prescriptive technical requirements contained in Option A of

Appendix J with performance based requirements and supporting

regulatory/industry documents contained in Option B of Appendix J.

This proposed change includes a description of the 10 CFR 50

Appendix J Testing Program Plan in Section 6.8.4.f of the Technical

Specifications.

This program plan, with one exception, is consistent with RG

[Regulatory Guide] 1.163. This exception to the RG is acceptable as

it is technically equivalent to and replaces an exemption that was

applicable to Option A of Appendix J. Therefore, this program plan

establishes leakage-rate test methods, procedures, acceptance

criteria and analyses which comply with Option B of Appendix J to 10

CFR 50.

The implementation of this program continues to provide adequate

assurance that during a DBA [Design Basis Accident]-LOCA the primary

containment and its components will continue to limit leakage rates

to less than the allowable leakage rates described in the Technical

Specifications and thereby limit leakage consistent with the

assumptions of the accident analyses. Therefore, the increased test

intervals permitted by Option B for the primary containment and its

penetrations will continue to implement the safety objectives

underlying the requirements of Appendix J.

As discussed under the margin of safety, the impact of the

proposed change on the consequences of a release is negligible. The

slight increase in the risk to the population is compensated by the

corresponding risk reduction benefits associated with the reduction

in component cycling, stress, and wear associated with increased

test intervals.

At least one air lock door in each air lock will continue to be

closed during the onset of an accident that would release

radioactivity into primary containment. Therefore, the air lock

interlocks continue to provide assurance that at least one leak

tested barrier will limit leakage during accident conditions.

The Hydrogen Recombiner System will continue to operate to

maintain the hydrogen concentration within the primary containment

below its flammable limit during post-LOCA conditions. This provides

assurance that primary containment integrity will not be challenged

by hydrogen burns.

Eliminating the inconsistency between the restoration statements

and the applicability requirements of LCO 3.6.1.2 does not diminish

the requirements contained in the Technical Specifications. The

Technical Specifications continue to require that the leakage limits

of LCO 3.6.1.2 be met prior to entering OPERATIONAL CONDITIONS 1, 2,

or 3 (i.e., temperature greater than 200 deg.F).

Accordingly, operation with the proposed change to the Operating

License, the Technical Specifications and the Bases will not

significantly increase the consequences of an accident previously

evaluated.

2. The operation of Nine Mile Point Unit 2, in accordance with

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

different kind of accident from any accident previously evaluated.

The proposed change would implement Option B of Appendix J of 10

CFR 50 for Type A, B and C testing. Option B would allow increased

testing intervals after satisfying certain performance based

criteria. The proposed change also corrects any inconsistency

between the restoration statements and the applicability

requirements of LCO 3.6.1.2. In addition, the proposed change

affects the testing intervals for the interlocks on the primary

containment air lock and for the measuring of the Hydrogen

Recombiner System leakage rate.

No new plant operating modes, system operating configurations

nor failure modes are introduced by the proposed change. The primary

containment and its penetrations will continue to perform their

accident mitigating function. The Hydrogen Recombiner System will

continue to function to prevent hydrogen burns within primary

containment during post-LOCA conditions.

Accordingly, operation with the proposed change will not create

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

accident previously evaluated.

3. The operation of Nine Mile Point Unit 2, in accordance with

the proposed amendment, will not involve a significant reduction in

a margin of safety.

A regulatory impact analysis of implementing performance-based

requirements indicates that relaxing the frequency of Type A, B and

C testing leads to an increase in overall reactor risk of

approximately two percent. As indicated in the Staff's Regulatory

Impact Analysis, this increase is considered to be marginal to

safety.

As indicated above, increasing test intervals can slightly

increase the risk to the population associated with the consequences

of a release; however, this is compensated by the corresponding risk

reduction benefits associated with the reduction in component

cycling, stress, and wear associated with increased test intervals.

Therefore, when considering the total integrated risk, the risk

associated with increased test intervals is negligible.

[[Page 20850]]

The proposed change is consistent with current plant safety

analyses. In addition, the proposed change does not require

revisions to the design of NMP2. As such, the proposed individual

changes will maintain the same level of reliability of the equipment

associated with containment integrity, assumed to operate in the

plant safety analysis, or provide continued assurance that specified

parameters affecting leak rate integrity, will remain within their

acceptance limits.

The as-left leakage after performing a required leakage test

continues to be less than 0.60 La for combined Type B and C leakage

and less than or equal to 0.75 La for Type A leakage. These as-left

acceptance criteria and the testing frequency as established by the

10 CFR 50 Appendix J Testing Program Plan provide assurance that the

measured leakage rate will not exceed the maximum allowable leakage

of La during plant operation.

Visual examination of accessible interior and exterior surfaces

of the primary containment continues to be performed prior to

initiating a Type A test. The total number of visual examinations

performed will continue to be three times during a 10-year period.

Therefore, visual examinations of the primary containment will

continue to allow for the timely uncovering of evidence of

structural deterioration and satisfy the requirements of RG 1.163.

The primary containment air lock interlocks will be tested prior

to conducting an air lock seal leakage test. This testing

requirement continues to provide adequate assurance that at least

one leak tested air lock door in each air lock will be closed during

accident conditions.

The measuring of the Hydrogen Recombiner System Leakage rate

will continue to be included as part of the overall integrated

leakage rate test. The test schedule for measuring system leakage

will also continue to coincide with the schedule for performing a

Type A test.

The leakage limits of LCO 3.6.1.2 will continue to be met prior

to entering into OPERATIONAL CONDITIONS 1, 2, or 3 (i.e.,

temperature greater than 200 deg.F). Satisfying these leakage

limits provides assurance that the measured leakage rate will not

exceed the maximum allowable leakage rate of La during plant

operation. Therefore, operation with 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Susan Frant Shankman, Acting

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: January 17, 1996.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) including revisions to

Specifications 3/4.3.1, ``Reactor Protection System Instrumentation,''

3/4.3.2, ``Isolation Actuation Instrumentation,'' 3/4.3.3, ``Emergency

Core Cooling System Actuation Instrumentation,'' 3/4.3.4.2, ``End-of-

Cycle Recirculation Pump Trip System Instrumentation,'' and the

associated Bases to relocate response time limit tables from the TSs to

the Updated Safety Analysis Report (USAR). The proposed revisions to

the TSs also include several administrative changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The operation of Nine Mile Point Unit 2, in accordance with

the proposed amendment, will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed amendment relocates Tables 3.3.1-2, ``Reactor

Protection System Response Times,'' 3.3.2-3, ``Isolation System

Instrumentation Response Times'' 3.3.3-3, ``Emergency Core Cooling

System Response Times'' and 3.3.4.2-3 ``End-of-Cycle Recirculation

Pump Trip System Response Time'' from the Technical Specifications

to the USAR. The Technical Specification Surveillance Requirements

and associated actions are not affected and remain in the Technical

Specifications. This change to the reactor protection system

instrumentation, isolation actuation instrumentation, and emergency

core cooling system instrumentation is being done in accordance with

the guidance provided in Generic Letter 93-08, ``Relocation of

Technical Specification Tables of Instrument Response Time Limits,''

and the change to the end-of-cycle recirculation pump trip system

instrumentation is consistent with NUREG 1433, ``Standard Technical

Specifications, BWR/4.'' This change allows NMP2 [Nine Mile Point

Unit 2] to administratively control subsequent changes to the

response time limits in accordance with 10CFR50.59. Additionally,

procedures which contain the various response time limits are also

subject to the change control provisions of 10 CFR 50.59. Relocating

this information does not affect the initial conditions of a design

basis accident or transient analysis. The proposed Technical

Specification changes do not affect the capability of the associated

systems to perform their intended functions within their required

response times. Since any subsequent changes to the USAR or

procedures which contain the response time limits are evaluated in

accordance with 10CFR50.59, the proposed amendment does not involve

an increase in the probability or consequences of an accident

previously evaluated.

2. The operation of Nine Mile Point Unit 2, in accordance with

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

different kind of accident from any previously evaluated.

The proposed change would relocate the response time limit

tables from the Technical Specifications to the USAR. Subsequent

changes to the USAR, or in procedures which contain the various

response time limits, would be evaluated in accordance with the

requirements of 10CFR50.59, which would evaluate the possibility of

the creation of a new or different kind of accident. The proposed

change does not involve any physical alteration of the plant, change

in a Limiting Condition for Operation or change in Surveillance

Requirements. No new failure modes are introduced. Therefore, this

proposed change does not create the possibility of a new or

different kind of accident from any previously evaluated.

3. The operation of Nine Mile Point Unit 2, in accordance with

the proposed amendment, will not involve a significant reduction in

a margin of safety.

The proposed change would relocate the response time limit

tables from the Technical Specifications to the USAR. Future changes

to the response time limits in the USAR, or in procedures which

contain the various response time limits, would be in accordance

with 10CFR50.59, which would evaluate the proposed change to

determine whether it involved any reduction in the margin of safety.

The response time limits to be transposed from the Technical

Specifications to the USAR are the same as the existing Technical

Specifications. Therefore, this proposed change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Susan Frant Shankman, Acting.

[[Page 20851]]

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: January 25, 1996.

Description of amendment request: The proposed amendment would

change a footnote in Table 3.3.3-1 and the corresponding footnote in

surveillance Table 4.3.3.1-1 (both referenced by Technical

Specification 3/4.3.3 ``Emergency Core Cooling System Actuation

Instrumentation'') to more clearly define when, during cold shutdown

and refueling (i.e., Operational Conditions 4 and 5), the Loss of

Voltage and Degraded Voltage relays associated with the 4.16 kV

Emergency Bus Undervoltage are required to be operable. The footnotes

currently state: ``Required when ESF [Engineered Safety Features]

equipment is required to be OPERABLE.'' The proposed amendment would

change the footnotes to state: ``Required when the associated diesel

generator is required to be OPERABLE.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequence of an accident previously evaluated.

The proposed change would require the Loss of Power instruments

to be OPERABLE in Operational Conditions 4 and 5 only when the

associated diesel generator is required to be OPERABLE. The Loss of

Power relays provide a support function to initiate the associated

diesel generator start and bus unloading sequences. If that diesel

generator is not in service, the loss of power relays perform no

safety function. Therefore, relating diesel generator OPERABILITY

and Loss of Power instrument OPERABILITY will not involve an

increase in the probability of an accident previously evaluated.

The proposed change does not affect the requirements of ESF

OPERABILITY. The change does not affect diesel generator response to

a loss of voltage or degraded voltage on the Divisional 4.16 kV

electrical busses when the diesel generator is required to be

OPERABLE. Automatic response of the ESF functions is unaffected by

removing the Loss of Power relays from service under these

conditions, therefore, the proposed change will not involve a

significant increase in the consequences of an accident previously

evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

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

different kind of accident from any accident previously evaluated.

The proposed change does not involve a modification of plant

equipment nor does it change the way the equipment will be

maintained or operated. The revision to Technical Specifications

will continue to require the Loss of Power instrumentation to be

OPERABLE when the associated diesel generator is required to be

OPERABLE. The Loss of Power instruments will continue to perform

their safety function of initiating the diesel generator start and

bus unloading sequences.

Therefore, this proposed change will not create the possibility

of a new or different kind of accident from any previously

evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The proposed change will not affect the OPERABILITY, operation

or reliability of any ESF function including the diesel generators.

All ESF functions will remain available during postulated accidents

with a loss of offsite electrical power. The change simply clarifies

when the Loss of Power instruments are required to be OPERABLE

during Operational Conditions 4 and 5. 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Susan Frant Shankman, Acting.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: March 15, 1996.

Description of amendment request: The proposed amendment would

revise the surveillance requirements of Technical Specification (TS)

4.6.2.1 ``Containment Systems--Depressurization Systems--Suppression

Pool'' to extend the time interval for performing the containment

drywell-to-suppression chamber bypass leakage test from 18 months to an

interval corresponding to that required for the Containment Integrated

Leak Rate Test. The provisions of TS 4.0.2 (which would provide an

extension of up to 25% of the specified surveillance interval) will not

apply. Specifically, existing TS 4.6.2.1.d would become subparagraphs d

and e to require that the suppression pool be demonstrated operable:

d. At least once per 18 months by conducting a visual inspection

of the exposed accessible interior and exterior surfaces of the

suppression chamber.*

e. At least every outage by requiring the performance of a

Containment Integrated Leak Rate Test, as scheduled in conformance

with the criteria specified in the 10 CFR 50 Appendix J Testing

Program Plan described in Section 6.8.4.f, by conducting a drywell-

to-suppression chamber bypass leak test at an initial differential

pressure of 3 psi and verifying that the [drywell-to-suppression

chamber bypass flow area] A/the square root of K calculated from the

measured leakage is within the specified limit of 0.0054 square

feet.

1. If any drywell-to-suppression chamber bypass leak test fails

to meet the specified limit, the test schedule for subsequent tests

shall be reviewed and approved by the Commission.

2. If two consecutive tests fail to meet the specified limit, a

test shall be performed at least each refueling outage until two

consecutive tests meet the specified limit, at which time the

original test schedule may be resumed.

3. The provisions of Specification 4.0.2 do not apply.

*Includes each vacuum relief valve and associated piping.

The proposed changes would also add a new surveillance requirement

for the testing of the bypass leakage path containing the suppression

chamber vacuum breakers, with associated acceptance criteria, which

would be performed each refueling outage that the bypass leak test is

not performed. Specifically, a new TS 4.6.2.1f would require that the

suppression pool be demonstrated operable:

f. During each refueling outage for which the drywell-to-

suppression chamber bypass leak test in Specification 4.6.2.1.e is

not conducted, by conducting a test of the four drywell-to-

suppression chamber bypass leak paths containing the suppression

chamber vacuum breakers at a differential pressure of at least 3 psi

and

1. Verifying that the total leakage area A/the square root of K

contributed by all four bypass leak paths is less than or equal to

24% of the specified limit, and

2. The leakage area for any one of the four bypass leak paths is

less than or equal to 12% of the specified limit.

By separate action, the NRC has provided notice of a proposed

amendment to change the frequency of Containment Integrated Leak Rate

Tests in accordance with Option B of 10 CFR Part 50 Appendix J . The

proposed changes described herein are intended

[[Page 20852]]

to be consistent with the changes proposed under Option B.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation on Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed TS changes involve the drywell-to-suppression

chamber bypass leak test frequency. There are no physical or

operational changes to the plant as a result of these proposed TS

revisions. Furthermore, the primary containment acts as an accident

mitigator and not as an accident initiator. Therefore, the proposed

TS changes do not affect the probability of any previously evaluated

accident.

The continued testing of bypass leakage pathways containing the

suppression chamber vacuum breakers on a refueling frequency, and

the continued requirement for visual inspection of containment

structural features assures that the bypass leakage path will not

degrade beyond the TS allowable limit during the interval between

performance of the bypass leakage test. Therefore, radioactivity

release following an accident will not be increased since the

pressure suppression capability of the containment is not reduced

from the existing design, and there will be no significant increase

in the consequences of any accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

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

different kind of accident from any accident previously evaluated.

The proposed 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 TS changes. These

proposed TS changes also include a requirement to continue

performing a surveillance test on the bypass leakage pathways

containing the vacuum breaker assemblies each refueling outage for

which the drywell-to-suppression chamber test is not conducted. This

test, along with the visual inspection required every refueling

cycle, will ensure that acceptable bypass leakage is maintained

during those intervals when the bypass leak test is not required.

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

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.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The drywell-to-suppression chamber bypass leak test data

obtained during previous testing at NMP2 [Nine Mile Point Unit 2]

demonstrates conformance, by a large margin, to the TS and design

leakage requirements. The test data and engineering evaluations

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 small risk of

bypass leakage through paths other than those containing the

suppression chamber vacuum breakers.

A test of the bypass leak pathways containing the vacuum

breakers will be used to verify acceptable bypass leakage during

those outages when the bypass leak test is not performed. The

proposed test of the bypass leak pathways containing the vacuum

breakers, 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. This

capability ensures that 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 [loss-

of-coolant accident] is maintained. Therefore, the proposed TS

changes do not involve a significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Susan Frant Shankman, Acting.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: March 20, 1996.

Description of amendment request: The proposed amendment would

revise Tables 3.3.1-1 and 4.3.1-1 of Technical Specification 3/4.3.1

``Reactor Protection System Instrumentation'' to delete the operability

requirement for the Average Power Range Monitor (APRM) Neutron Flux-

Upscale, Setdown and Inoperative functions in Operational Conditions

(OCs) 3 (Hot Shutdown) and 4 (Cold Shutdown). These same functions

would also be revised for OC 5 (Refueling) to indicate that operability

will only be required during shutdown margin demonstrations performed

per TS 3.10.3.

Basis for proposed no significant hazards consideration

determination: The revisions to the APRM functions are proposed to

support licensee's plans to replace Local Power Range Monitors during

the next refueling outage. The revisions also provide for the eventual

replacement of the existing APRM System with the Nuclear Measurement

Analysis and Control Power Range Neutron Monitoring System, and the

eventual installation of the Oscillation Power Range Monitor system for

the detection of reactor instability conditions. These modifications

are based upon Report NEDO-31960, ``BWR Owners' Group Long-Term

Solutions Licensing Methodology, approved by the Commission July 12,

1993; the licensee's response of November 8, 1994, selecting Option III

in NEDO-31960 for Nine Mile Point, Unit 2; NRC Generic Letter 94-02,

``Long-Term Solutions and Upgrade of Interim Operating Recommendations

for Thermal-Hydraulic Instabilities in Boiling Water Reactors'' dated

July 11, 1994; and General Electric Licensing Topical Report, NEDC-

32410P-A, ``Nuclear Measurement Analysis and Control Power Range

Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip

Function,'' which was approved by the Commission September 5, 1995.

As required by 10 CFR 50.91(a), the licensee has provided its

analysis of the issue of no significant hazards consideration, which is

presented below:

The operation of Nine Mile Point Unit 2 in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The Reactor Protection System (RPS) initiates a reactor scram

when one or more monitored parameters exceed their specified limits

to preserve the integrity of the fuel cladding and the Reactor

Coolant System and to minimize the energy that must be absorbed

following a loss-of-coolant accident. The proposed changes will

revise the OCs in which the APRM Neutron Flux-Upscale, Setdown and

Inoperative RPS Instrumentation is required. These changes do not

affect the probability of precursors of any accidents previously

evaluated, and therefore, do not increase their probability.

During normal operation in OCs 3 and 4, all control rods are

fully inserted and the reactor mode switch position control rod

withdrawal blocks do not allow control rods to be withdrawn.

Therefore, the RPS APRM functions are not required. Specification

3.9.10 does allow one control rod to be removed from the core in OC

4 by placing the mode switch in the refuel position. However, with

the reactor mode switch in the refuel position, refueling interlocks

are in place (i.e., one-rod out, etc.), which together with

[[Page 20853]]

adequate shutdown margin will preclude unacceptable reactivity

excursions. The APRM Neutron Flux-Upscale, Setdown function is not

required during OC 5 except during shutdown margin demonstrations.

The SRMs [source range monitors], IRMs [intermediate range

monitors], and refueling interlocks provide adequate protection from

reactivity excursions during OC 5. The exception is during the

shutdown margin demonstration when more than one control rod will be

withdrawn and the APRMs will continue to be required to be operable

as a backup to the IRMs. Testing of the RPS APRM functions will

continue to be performed in those OCs for which operability is

required. Consequently, the reliability and performance of the RPS

APRM functions in these OCs will not be adversely affected.

Therefore, the proposed change will not result in a significant

increase in the consequences of any accidents previously evaluated.

The operation of Nine Mile Point Unit 2 in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes will revise the applicable OCs in which the

APRM neutron Flux-Upscale, Setdown and Inoperative RPS

instrumentation is required. Changes to OC requirements will not

introduce any new accident precursors and will not involve any

physical alternations to plant configurations which could initiate a

new or different kind of accident. NMP2 is analyzed for a single

control rod withdrawal error during refueling. Since the core is

designed to meet shutdown requirements with the highest worth rod

withdrawn, the core remains subcritical even with one rod withdrawn.

The one-rod-out interlock which allows only one control rod to be

withdrawn in OC 5 is not affected by the proposed changes.

Consequently, the proposed changes do not create an accident

different than the previously analyzed single control rod withdrawal

error event. Surveillance testing will continue to be performed to

assure reliability and maintain current performance levels.

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

new or different kind of accident from any previously evaluated.

The operation of Nine Mile Point Unit 2 in accordance with the

proposed amendment will not involve a significant reduction in a

margin of safety.

The proposed changes to the RPS APRM function instrumentation

Technical Specification requirements will not adversely affect the

design or the performance characteristics of the RPS instrumentation

nor will it affect the ability of the RPS APRM instrumentation to

perform its intended function. As discussed above, the subject RPS

instrumentation is not required in OC 3, 4, and 5 except for

shutdown margin demonstrations. Accordingly, deletion of the

requirement to have these functions operable in these OCs will not

significantly reduce a margin of safety. Surveillance testing will

continue to be performed for those OCs in which the instrumentation

is required to assure reliability. Therefore, the proposed changes

do not involve a significant reduction in a margin of safety

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Susan Frant Shankman, Acting.

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

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

Date of amendment request: March 28, 1996.

Description of amendment request: The proposed amendment would

change Technical Specification Section 3.7.7, ``Sealed Source

Contamination,'' by making the criteria for testing sealed sources for

contamination and leakage at Millstone Unit No. 2 the same as those at

Millstone Unit No. 3, the Haddam Neck Plant, and Seabrook Station.

Specifically, the sealed sources that are required to be free of

greater than or equal to 0.005 microcuries of removable contamination

would be those that would exceed ``100 microcuries of beta and/or gamma

emitting material or 5 microcuries of alpha emitting material.'' The

Bases Section 3/4.7.7, ``Sealed Source Contamination,'' would also be

changed to reference the appropriate section of 10 CFR 70.39.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (SHC), which is presented below:

Pursuant to 10 CFR 50.92, NNECO [Northeast Nuclear Energy

Company] has reviewed the proposed changes and concludes that the

changes do not involve a significant hazards consideration (SHC)

since the proposed changes satisfy the criteria in 10 CFR 50.92(c).

That is, the proposed changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The changes make the criteria for testing sealed sources for

contamination and leakage at Millstone Unit No. 2 the same as those

at Millstone Unit No. 3, the Haddam Neck Plant and Seabrook Station.

Although the leakage criteria for sealed sources that are to be

tested is being changed, the allowable leakage remains small. Any

leakage that is identified would not cause a significant radiation

exposure. The source storage area is routinely surveyed by Health

Physics in accordance with Health Physics Department procedures and

any significant leakage would be detected. Therefore, this change

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

from any accident previously evaluated.

The proposed change in the criteria for testing sealed sources

for contamination and leakage will not change the way the sources

are used. Therefore, this change will not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

The possible radiation exposure to both the workers and the

public from this change is very small. All protective systems which

would detect any release of material from the site remain in place

so there is no reduction in safety for the public. Likewise, all

protective systems for the workers remain in place. Workers using

the sources routinely pass through the whole body contamination

monitors. In addition, the source storage areas are surveyed

routinely by Health Physics in accordance with Health Physics

Department procedures, and any significant leakage would be

detected. The bases section is being revised to reference the

appropriate section of 10 CFR 70.39. Therefore, there is no

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360, and Waterford Library, Attn: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385.

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270.

NRC Project Director: Phillip F. McKee.

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: March 12, 1996.

[[Page 20854]]

Description of amendment request: The proposed changes would remove

a requirement to interconnect two or more accumulators for the purpose

of cross checking instrumentation in the event that one of the two

pressure or level instrument channels on an accumulator is declared

inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response: The design basis accident for which the accumulators

were designed is the double ended guillotine break of a cold leg.

Interconnecting or not interconnecting accumulators does not have

any effect on the probability of occurrence of this event. By

eliminating the requirement to interconnect accumulators, the

proposed amendment assures that a minimum of three accumulators are

available, as assumed in the safety analyses, to mitigate the

consequences of a large-break loss-of-coolant [LBLOCA] accident.

Therefore, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously analyzed.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response: The proposed amendment does not involve any physical

changes to plant equipment or setpoints and does not create the

possibility of a new or different kind of accident. Eliminating the

requirement to interconnect accumulators ensures that the plant

configuration is maintained consistent with that assumed in the

safety analysis and no new failure modes are created.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response: There is no margin of safety specified in the

Technical Specifications for these instrument channels. There are no

setpoints or allowable values associated with these instrument

channels which affect Safety Limits or Limiting Safety System

Settings. The proposed amendment ensures that the safety analysis

assumption regarding the accumulators remains valid and the

resulting peak fuel clad temperature meets specified acceptance

criteria. The proposed amendment does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,

New York, New York 10019.

NRC Project Director: Susan Frant Shankman, Acting.

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: March 14, 1996.

Description of amendment request: The proposed changes would allow

a one-time extension of the inspection interval for the steam generator

tubes that is due in July 1996.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response: The proposed license amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated. As stated in the Basis of the IP3

[Indian Point Unit 3] Technical Specifications, the program for

inservice inspection of steam generator tubes regarding equipment,

procedures, and sample selection is based upon the guidance and

recommendations in Regulatory Guide 1.83 and NRC Generic Letter 85-

02. The addition of the footnote to extend the surveillance due date

will not increase the deviation from the guidance and recommendation

stated above, and, therefore will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response: The proposed license amendment does not create the

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

previously evaluated. The proposed change does not involve the

addition of any new or different type of equipment, nor does it

involve the operation of equipment required for safe operation of

the facility in a manner different from those addressed in the Final

Safety Analysis Report. Therefore, the proposed change will not

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

any accident previously evaluated.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response: The proposed license amendment does not involve a

significant reduction in a margin of safety. The proposed change

does not adversely affect any safety related system or component

operation or operability, instrument operation, or safety system

setpoints and does not result in increased severity of any of the

accidents considered in the safety analysis. This change has no

adverse effect on any margin of safety and, therefore, does not

create a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,

New York, New York 10019.

NRC Project Director: Susan Frant Shankman, Acting

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: March 22, 1996.

Description of amendment request: The amendment proposes changes to

the Technical Specifications (TS) to establish operability requirements

for avoidance and protection from thermal hydraulic instabilities to be

consistent with Boiling Water Reactor Owners Group long-term solution

Option I-D. Editorial changes are also made to support the revised

specifications, improve readability of Bases sections, and enhance the

presentation of requirements for single loop operation.

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:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92, since it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The implementation of BWR Owners' Group long-term stability

solution Option I-D at FitzPatrick does not modify the assumptions

contained in the existing accident analysis. The use of an exclusion

region and the operator actions required to avoid and minimize

operation inside the region do not increase the possibility of an

accident. Conditions of operation outside of the exclusion region

are within the analytical envelope of the existing safety analysis.

The operator action requirement to exit the

[[Page 20855]]

exclusion region upon entry minimizes the possibility of an

oscillation occurring. The actions to drive control rods and/or to

increase recirculation flow to exit the region are maneuvers within

the envelope of normal plant evolutions. The flow referenced scram

has been analyzed and will provide automatic fuel protection in the

event of an instability. Thus, each proposed operating requirement

provides defense in depth for protection from an instability event

while maintaining the existing assumptions of the accident analysis.

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

from those previously evaluated because:

The proposed operating requirements either mandate operation

within the envelope of existing plant operating conditions or force

specific operating maneuvers within those carried out in normal

operation. Since operation of the plant with all of the proposed

requirements are within the existing operating basis, an unanalyzed

accident will not be created through implementation of the proposed

change.

3. Involve a significant reduction in the margin of safety

because:

Each of the proposed requirements for plant thermal hydraulic

stability provides a means for fuel protection. The combination of

avoiding possible unstable conditions and the automatic flow

referenced reactor scram provides an in depth means for fuel

protection. Therefore, the individual or combination of means to

avoid and suppress an instability supplements the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Susan Frank Shankman, Acting.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: March 22, 1996.

Description of amendment request: The amendment proposes to revise

Technical Specification (TS) Table 3.2-2, ``Core and Containment

Cooling System Initiation and Control Instrumentation Operability

Requirements.'' The proposed changes will revise allowed outage times

(AOTs) for 4kV Emergency Bus Undervoltage Trip Functions. The AOTs for

these trip functions were extended by Amendment 227; however, the AOT

extensions for these trip functions were not consistent with the

requirements of Standard Technical Specifications (STS), NUREG-1433,

and differed from the recommendations in the associated Licensing

Topical Report. Additional changes are proposed to TS Table 3.2-2 and

to TS Table 4.2-2, ``Core and Containment Cooling System

Instrumentation Test and Calibration Requirements.'' These changes

will: (1) replace the generic actions for inoperable instrument

channels with function-specific actions, (2) replace the generic test

AOT with function-specific test AOTs, and (3) relocate selected trip

functions from the TS to an Authority controlled document.

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:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92, since it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes are limited to replacement of the generic

actions and test AOT with function-specific actions and test AOTs,

and relocation of selected trip functions from the TS to an

Authority controlled document. The changes do not introduce any new

modes of plant operation, make any physical changes, or alter any

operational setpoints. Therefore, the changes do not degrade the

performance of any safety system assumed to function in the accident

analysis. Consequently, there is no effect on the probability or

consequences of an accident.

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

from those previously evaluated.

The proposed changes do not introduce any new accident

initiators or failure mechanisms since the changes do not introduce

any new modes of plant operation, make any physical changes, or

alter any operational setpoints. Therefore the changes do not create

the possibility of a new or different kind of accident.

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

The proposed changes do not alter the manner in which safety

limits, limiting safety system settings, or limiting conditions for

operation are determined. The relocated requirements do not satisfy

the 10 CFR 50.36 criteria for inclusion in the Technical

Specifications. Therefore, the changes do not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Susan Frant Shankman, Acting.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: March 27, 1996.

Description of amendment request: The amendment proposes to revise

the Technical Specifications to support adoption of the primary

containment leakage rate testing requirements of Option B to 10 CFR 50,

Appendix J at the FitzPatrick plant, and clarify the numerical value of

the allowable containment leakage rate (La) as 1.5 percent per

day.

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 Authority has evaluated the proposed TS Amendment and

determined that it does not represent a significant hazards

consideration. Based on the criteria for defining a significant

hazards consideration established in 10 CFR 50.92, operation of the

James A. FitzPatrick Nuclear Power Plant in accordance with the

proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The proposed changes do not involve a change to the design or

operation of the plant. The systems affected by this proposed TS

change are not assumed in any safety analyses to initiate any

accident sequence. Therefore, the probability of any accident

previously evaluated is not increased by this proposed TS change.

The clarification of the allowable containment leakage rate

(La) is consistent with the accident analyses. There is no

change to the consequences of an accident previously evaluated

because maintaining leakage within limits assumed in the accident

analyses ensures that the dose consequences resulting from an

accident are not increased. The proposed TS changes maintain an

equivalent level of reliability

[[Page 20856]]

and availability for all affected systems. The ability of the

affected systems associated with maintaining leak rate integrity to

perform their intended function is unaffected by the proposed TS

changes. Implementation of these changes will provide continued

assurance that specified parameters associated with containment

integrity will remain within acceptance limits, and as such, will

not significantly increase the consequences of a previously

evaluated accident.

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

from any accident previously evaluated because:

The proposed changes allow adoption of those requirements

specified in Option B to 10 CFR 50, Appendix J, and do not involve a

change to the plant design and operation. As a result, the proposed

changes do not affect the parameters or conditions that could

contribute to the initiation of any accidents. The methods of

performing primary containment leakage rate testing are not changed.

No new accident modes are created by allowing extended intervals for

Type A, B and C testing, or by clarifying the numerical value of the

allowable containment leakage rate (La). No safety-related

equipment or safety functions are altered, or adversely affected, as

a result of these changes. The proposed changes will not introduce

failure mechanisms beyond those already considered in the current

plant safety analyses. Extension of the test intervals, and

clarification of the allowable leakage rate, does not contribute to

the possibility of a new or different kind of accident or

malfunction from those previously analyzed.

3. Involve a significant reduction in the margin of safety

because: The proposed changes affect the frequency of primary

containment leakage rate testing, and the numerical definition of

the allowable containment leakage rate (La). The design of the

FitzPatrick plant is not changed. The methodology for test

performance is unchanged and Type A, B and C tests will continue to

be performed at Pa. The proposed changes provide

sufficient controls to ensure that proper maintenance and repairs

are performed on the primary containment, and systems and components

penetrating the primary containment. The reliability of containment

systems assumed to operate in the plant safety analyses is not

reduced. The numerical value of La specified in Specification

6.20 is consistent with the accident analyses, therefore, the dose

consequences of any analyzed accidents are not increased. Therefore,

the proposed changes provide continued assurance of the leak

tightness of the containment without adversely affecting the public

health and safety and, as such, will not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Susan Frant Shankman, Acting.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of amendment request: April 22, 1996.

Description of amendment request: The amendments would change the

Technical Specifications to implement 10 CFR Part 50, Appendix J,

Option B, for the Type A test by referring to Regulatory Guide 1.163,

``Performance-Based Containment Leakage-Test Program.''

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.

Containment leak rate testing is not an initiator of any

accident. The proposed changes do not make any physical changes to

the containment. The proposed changes do not affect performance of

the containment, reactor operations or accident analysis. Therefore,

the proposed changes will not involve an increase in the probability

of any previously evaluated accident.

Since the allowable leakage rate is not being changed and since

the analysis documented in NUREG-1493, ``Performance-Based

Containment Leak-Test Program'' concludes that the impact on public

health and safety due to extended intervals is negligible, the

proposed changes will not involve an increase in the consequences of

any previously evaluated accident. Therefore, adoption of a

performance-based verification of leakage rates for the overall

containment boundary will provide an equivalent level of safety and

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change makes no physical changes to the plant.

Since no physical changes are involved and since the analysis

documented in NUREG-1493 confirms that the performance based

schedule continues to maintain a minimal impact on public risk, it

can be concluded that the effect of the containment on any accident

will not change. The proposed change does not affect normal plant

operations or configuration, nor does it affect leak rate test

pressure.

Therefore, the proposed amendment will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes are based on NRC-accepted provisions, and

maintain necessary levels of reliability of containment integrity.

The performance-based approach to leakage rate testing recognizes

that historically good results of containment testing provide

appropriate assurance of future containment integrity. This supports

the conclusion that the impact on the health and safety of the

public as a result of extended test intervals is negligible. Since

the analysis documented in NUREG-1493 confirms that the performance

based schedule continues to maintain a minimal impact on public

risk, it can be concluded that the margin of safety is not

significantly affected by the proposed changes.

The test history at Salem Units 1 and 2 (no ILRT failures)

provides continued assurance of the leak tightness of the

containment structure.

Therefore, the proposed amendment will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Salem Free Public library, 112

West Broadway, Salem, New Jersey 08079.

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

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

NRC Project Director: John F. Stolz.

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

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

Date of amendment request: April 4, 1996 (TS 96-01).

Description of amendment request: The proposed change would revise

the appropriate technical specifications, surveillances, and bases as

needed for the conversion from Westinghouse nuclear fuel to Framatome

Cogema Mark-BW17 nuclear fuel.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

[[Page 20857]]

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

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

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The analyses provided in Topical Report BAW-10220P show that the

changes do not significantly change the results of previously

evaluated events. These analyses provide the template for accident

analyses assumptions that must be met by the cycle-specific reload

analysis.

The SQN Units 1 and 2 Cycle 9 reload cores with Mark-BW fuel

will be designed to operate within the approved limits for accident

analysis. The limits provided in the TS and described in the Updated

Final Safety Analysis Report (UFSAR) provide the framework for

accident analyses. By maintaining these limits, the probability or

consequences of accidents related to the core changes do not

significantly change. Thus, it is concluded that there is no

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any previously analyzed.

The change to Mark-BW fuel cores and mixed (transition) cores

has been evaluated in the Topical Report BAW-10220P. It was

concluded that the change did not create new or different kinds of

accidents. The change in fuel suppliers has been evaluated for

consideration of the effects of power distribution and peaking

factors such that there are no restrictions on the use of Mark-BW

fuel assemblies beyond those already established in the UFSAR and

TS. Adherence to the safety analysis limits restricts the

possibility of new or different accidents. Historically, new

accidents have not been associated with changes in fuel suppliers as

long as safety analysis limits continue to be met. It is concluded

that transition to Mark-BW fuel does not create the possibility of a

new or different kind of accident from any previously analyzed.

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

The margin of safety is established by the acceptance criteria

used by NRC. Meeting the acceptance criteria assures that the

consequences of accidents are within known and acceptable limits.

The loss-of-coolant accident (LOCA) acceptance criteria are

unchanged: peak cladding temperature of 2200 degrees

Fahrenheit, peak cladding oxidation of 17 percent,

average clad oxidation of 1 percent, and long-term

coolability. These requirements continue to be met. The methods used

to demonstrate conformance with these limits have changed, and were

reviewed to assure that the methods, as well as the results, are

acceptable. The acceptance criteria for Departure from Nucleate

Boiling (DNB) events has not changed and is still the 95 percent

probability and 95 percent confidence interval that DNB is not

occurring during the transient. The DNB correlation, and methods

used to demonstrate that DNB limits are met, have changed, and these

changes were reviewed to assure conformance with acceptable

practices. Other changes, as well as the changes discussed above,

have been evaluated in the referenced safety analyses and are shown

to meet applicable acceptance criteria. Other margins, such as

avoiding fuel centerline melting are not significantly changed.

Based on these results, it is concluded that the margin of safety is

not significantly reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon.

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

Callaway County, Missouri

Date of application request: April 12, 1996.

Description of amendment request: The proposed amendment would

change Technical Specification (TS) 3/4.4 and its associated Bases to

address the installation of laser welded tube sleeves in the Callaway

Plant steam generators.

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 elevated tubesheet LWS [laser welded sleeve] configuration

has been designed and analyzed in accordance with the requirements

of the ASME [American Society of Mechanical Engineers] Code. The

applied stresses and fatigue usage for the sleeve and weld are

bounded by the limits established in the ASME Code. ASME Code

minimum material property values are used for the structural and

plugging limit analysis. Ultrasonic inspection is used to verify

that minimum weld fusion zone thickness are produced. Mechanical

testing has shown that the individual joint structural strength of

Alloy 690 LWS under normal, upset and faulted conditions provides

margin to the acceptance limits. These acceptance limits bound the

most limiting (3 times normal operating pressure differential) burst

margin recommended by RG [Regulatory Guide] 1.121. Therefore, each

individual joint provides for structural integrity exceeding RG

recommendations.

Leakage testing for \7/8\'' and \3/4\'' tube sleeves has

demonstrated that no unacceptable levels of primary to secondary

leakage are expected during any plant condition, including the case

where the seal weld is not produced in the lower joint of the

tubesheet sleeve. Similar tests of 11/16'' tube sleeves will be

completed prior to Refuel 8.

The sleeve minimum acceptable wall thickness (used for

developing the depth-based plugging limit for the sleeve) is

determined using the guidance of Regulatory Guide 1.121 and the

pressure stress equation of Section III of the ASME Code. The

limiting requirement of Regulatory Guide 1.121, which applies to

part throughwall degradation, is that the minimum acceptable wall

must maintain a factor of safety of three against tube failure under

normal operating (design) conditions. A bounding set of design and

transient loading input conditions was used for the minimum wall

thickness evaluation in the generic evaluation. Evaluation of the

minimum acceptable wall thickness for normal, upset and postulated

accident condition loading per the ASME Code indicates these

conditions are bounded by the design condition requirement minimum

wall thickness.

A bounding tube wall degradation growth rate per cycle and an

eddy current uncertainty has been assumed for determining the sleeve

TS plugging limit. The sleeve wall degradation extent determined by

eddy current examination, which would require plugging sleeved

tubes, is developed using the guidance of RG 1.121 and is defined in

WCAP-14596 to be 39 percent throughwall of the sleeve nominal wall

thickness.

The consequences of failure of the sleeve joint are bounded by

the current steam generator tube rupture analysis included in the

Callaway FSAR. Due to the slight reduction in diameter caused by the

sleeve wall thickness, primary coolant release rates would be

slightly less than assumed for the steam generator tube rupture

analysis (depending on the break location), and therefore, would

result in lower total primary fluid mass release to the secondary

system.

The proposed change does not adversely impact any other

previously evaluated design basis accident of the results of LOCA

and non-LOCA accident analyses for the current TS minimum reactor

coolant system flow rate. The results of the analyses and testing

demonstrate that the sleeve assembly is an acceptable means of

maintaining tube integrity. Furthermore, per Regulatory Guide 1.83,

``Inservice Inspection of Pressurized Water Reactor Steam Generator

Tubes'' recommendations, the sleeved tube can be monitored through

periodic inspections with present eddy current techniques. These

measures demonstrate that installation of sleeves spanning degraded

areas of the tube will restore the tube to a condition consistent

with its original design basis.

Corrosion testing of laser welded sleeve joints indicates that

the corrosion resistance

[[Page 20858]]

(relative to roll transition control samples) can be increased by

greater than a factor of ten with the application of a post weld

heat treatment [PWHT]. All free span laser welds will receive a post

weld heat treatment. Therefore, rapid corrosion degradation of the

free span laser weld joint region is not expected. Recently

performed corrosion testing of LWS joints in locked (at the first

TSP [tube support plate] structure) tube conditions indicates that

the PWHT, the stress corrosion cracking initiation potential in the

weld region of the parent tube is reduced and the cracking

resistance is enhanced. Similar test results and conclusions would

be expected for Callaway based on the similarity of designs and

expected tube far field residual stresses.

Conformance of the sleeve design with the applicable sections of

the ASME Code and results of the leakage and mechanical tests,

support the conclusion that installation of LWS will not increase

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.

Sleeving will not adversely affect any plant component. Stress

and fatigue analysis of the repair has shown that the ASME Code and

Regulatory Guide 1.121 criteria are not exceeded. Implementation of

LWS maintains overall tube bundle structural and leakage integrity

at a level consistent to that of the originally supplied tubing

during all plant conditions. Leak and mechanical testing of sleeves

support the conclusions of the calculations that each sleeve joint

retains both structural and leakage integrity during all conditions.

Sleeving of tubes does not provide a mechanism resulting in an

accident outside of the area affected by the sleeves. Any accident

as a result of potential tube or sleeve degradation in the repaired

portion of the tube is bounded by the existing tube rupture accident

analysis.

Implementation of LWS will reduce the potential for primary to

secondary leakage during a postulated steam line break while not

significantly impacting available primary coolant flow area in the

event of a LOCA. By effectively isolating degraded areas of the tube

through repair, the potential for steam line break leakage is

reduced. These degraded intersections now are returned to a

condition consistent with the Design Basis. While the installation

of a sleeve reduces primary coolant flow, the reduction is far below

that caused by plugging. Therefore, far greater primary coolant flow

area is maintained through sleeving versus plugging.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The LWS repair of degraded steam generator tubes has been shown

by analysis to restore the integrity of the tube bundle consistent

with its original design basis condition, i.e., tube/sleeve

operational and faulted condition stresses are bounded by the ASME

Code requirements and the repaired tubes are leaktight. The safety

factors used in the design of sleeves for the repair of degraded

tubes are consistent with the safety factors in the ASME Code used

in steam generator design. The design of the tubesheet sleeve lower

joints for the \3/4\'' and \7/8\'' sleeves have been verified by

testing to preclude leakage during normal and postulated accident

conditions. Similar tests of \11/16\'' sleeves will be completed

prior to Refuel 8. The portions of the installed sleeve assembly

which represent the reactor coolant pressure boundary can be

monitored for the initiation and progression of sleeve/tube wall

degradation, thus satisfying the requirements of Regulatory Guide

1.83. The portion of the tube bridged by the sleeve joints is

effectively removed from the pressure boundary, and the sleeve then

forms the new pressure boundary. The areas of the sleeved tube

assembly which require inspection are defined in WCAP-14596.

In addition, since the installed sleeve represents a portion of

the pressure boundary, a baseline inspection of these areas is

required prior to operation with sleeves installed. The effect of

sleeving on the design transients and accident analyses has been

reviewed based on the installation of sleeves up to the level of

steam generator tube plugging coincident with the minimum reactor

flow rate and the Callaway Safety Analysis.

Provisional requirements cited in other NRC Safety Evaluation

Reports addressing the implementation of sleeving have required the

reduction of the individual steam generator normal operation primary

to secondary leakage limit from 500 to 150 gpd. Consistent with

these evaluations, Union Electric will reduce the per steam

generator leak rate limit of 500 gpd in TS 3.4.6.2.c to 150 gpd. The

establishment of this leakage limit at 150 gpd provides additional

safety margin.

Finally, Union Electric will reduce the tube plugging limit from

48 percent through wall to 40 percent through wall to be consistent

with NUREG-1431. The establishment of the plugging limit at 40

percent through wall provides additional safety margin.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

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

Callaway County, Missouri

Date of application request: April 12, 1996.

Description of amendment request: The proposed amendment would

change Technical Specification (TS) 3/4.4 and its associated Bases to

address the installation of electrosleeves in the Callaway Plant steam

generators.

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 electrosleeve configuration has been designed and analyzed

in accordance with the requirements of the ASME [American Society of

Mechanical Engineers] Code. The applied stresses and fatigue usage

for the sleeve are bounded by the limits established in the ASME

Code. ASME Code minimum material property values are used for the

structural and plugging limit analysis. Mechanical testing has shown

that the structural strength of nickel electrosleeves under normal,

upset and faulted conditions provides margin to the acceptance

limits. These acceptance limits bound the most limiting (3 times

normal operating pressure differential) burst margin recommended by

RG [Regulatory Guide] 1.121. Leakage testing for \5/8\'', \7/8\''

and \3/4\'' tube sleeves has demonstrated that no unacceptable

levels of primary to secondary leakage are expected during any plant

condition. Similar tests of \11/16\'' tube electrosleeves will be

completed prior to Refuel 8.

The sleeve nominal wall thickness (used for developing the

depth-based plugging limit for the sleeve) is determined using the

guidance of Regulatory Guide 1.121 and the pressure stress equation

of Section III of the ASME Code. The limiting requirement of

Regulatory Guide 1.121, which applies to part throughwall

degradation, is that the minimum acceptable wall must maintain a

factor of safety of three against tube failure under normal

operating (design) conditions. A bounding set of design and

transient loading input conditions was used for the minimum wall

thickness evaluation in the generic evaluation. Evaluation of the

minimum acceptable wall thickness for normal, upset and postulated

accident condition loading per the ASME Code indicates these

conditions are bounded by the design condition requirement minimum

wall thickness.

A bounding tube wall degradation growth rate per cycle and an

NDE [nondestructive examination] uncertainty has been assumed for

determining the sleeve TS plugging limit. The sleeve wall

degradation extent determined by NDE, which would require plugging

sleeved tubes, is developed using the guidance of RG 1.121 and is

defined in BAW-10219P to be 20 percent throughwall.

The consequences of failure of the sleeve are bounded by the

current steam generator tube rupture analysis included in the

Callaway FSAR [final safety analysis report]. Due to the slight

reduction in diameter caused by the sleeve wall thickness, primary

coolant release rates would be slightly less

[[Page 20859]]

than assumed for the steam generator tube rupture analysis

(depending on the break location), and therefore, would result in

lower total primary fluid mass release to the secondary system.

The proposed change does not adversely impact any other

previously evaluated design basis accident or the results of LOCA

[loss-of-coolant accident] and non-LOCA accident analyses for the

current TS minimum reactor coolant system flow rate. The results of

the analyses and testing demonstrate that the electrosleeve is an

acceptable means of maintaining tube integrity. Furthermore, per

Regulatory Guide 1.83 recommendations, the sleeved tube can be

monitored through periodic inspections with present NDE techniques.

These measures demonstrate that installation of sleeves spanning

degraded areas of the tube will restore the tube to a condition

consistent with its original design basis.

Conformance of the electrosleeve design with the applicable

sections of the ASME Code and results of the leakage and mechanical

tests, support the conclusion that installation of electrosleeves

will 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.

Electrosleeving does not represent a potential to adversely

affect any plant component. Stress and fatigue analysis of the

repair has shown that the ASME Code and Regulatory Guide 1.121

criteria are not exceeded. Implementation of electrosleeving

maintains overall tube bundle structural and leakage integrity at a

level consistent to that of the originally supplied tubing during

all plant conditions. Leak and mechanical testing of electrosleeves

support the conclusions of the calculations that each sleeve retains

both structural and leakage integrity during all conditions.

Sleeving of tubes does not provide a mechanism resulting in an

accident outside of the area affected by the sleeves. Any accident

as a result of potential tube or sleeve degradation in the repaired

portion of the tube is bounded by the existing tube rupture accident

analysis.

Implementation of sleeving will reduce the potential for primary

to secondary leakage during a postulated steam line break while not

significantly impacting available primary coolant flow area in the

event of a LOCA. By effectively isolating degraded areas of the tube

through repair, the potential for steam line break leakage is

reduced. These degraded intersections now are returned to a

condition consistent with the Design Basis. While the installation

of a sleeve reduces primary coolant flow, the reduction is far below

that caused by plugging. Therefore, far greater primary coolant flow

area is maintained through sleeving versus plugging.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The electrosleeve repair of degraded steam generator tubes has

been shown by analysis to restore the integrity of the tube bundle

consistent with its original design basis condition, i.e., tube/

sleeve operational and faulted condition stresses are bounded by the

ASME Code requirements and the repaired tubes are leaktight. The

safety factors used in the design of sleeves for the repair of

degraded tubes are consistent with the safety factors in the ASME

Code used in steam generator design. The portions of the installed

sleeve assembly which represent the reactor coolant pressure

boundary can be monitored for the initiation and progression of

sleeve/tube wall degradation, thus satisfying the requirements of

Regulatory Guide 1.83. The portion of the tube bridged by the sleeve

is effectively removed from the pressure boundary, and the sleeve

then forms the new pressure boundary. The areas of the sleeved tube

assembly which require inspection are defined in BAW-10219P.

In addition, since the installed sleeve represents a portion of

the pressure boundary, a baseline inspection of these areas is

required prior to operation with sleeves installed. The effect of

sleeving on the design transients and accident analyses has been

reviewed based on the installation of sleeves up to the level of

steam generator tube plugging coincident with the minimum reactor

flow rate and the Callaway Safety Analysis.

Provisional requirements cited in other NRC Safety Evaluation

Reports addressing the implementation of sleeving have required the

reduction of the individual steam generator normal operation primary

to secondary leakage limit from 500 to 150 gpd.

Consistent with these evaluations, Union Electric will reduce

the per steam generator leak rate limit of 500 gpd in TS 3.4.6.2.c

to 150 gpd. The establishment of this leakage limit at 150 gpd

provides additional safety margin.

Finally, Union Electric will reduce the tube plugging limit from

48 percent through wall to 40 percent through wall to be consistent

with NUREG-1431. The establishment of the plugging limit at 40

percent through wall provides additional safety margin.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: April 4, 1996.

Description of amendment request: The proposed amendment would

revise the Technical Specifications regarding secondary containment

integrity including addition of required actions in the event secondary

containment integrity is not maintained when required. It would also

require surveillance of the secondary containment isolation valves

under the licensee's in-service testing program.

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 result in any hardware changes.

The requirements for Secondary Containment integrity are not assumed

in the initiation of any analyzed event. The proposed changes

establish and maintain adequate assurance that Secondary Containment

Integrity will be maintained as assumed in analyses for the

mitigation of accident consequences. Not requiring Secondary

Containment Integrity when the reactor coolant system is not vented

in the Cold Shutdown condition or the Refuel Mode does not involve

an increase in previously evaluated accident consequences since no

mechanism exists to impart additional fission-products into the

reactor coolant. Under these conditions, activities for which the

reactor coolant system would not be vented would be strictly

controlled and monitored. As a result, leaks or pipe breaks would

typically be detected before significant inventory loss occurred.

These activities would typically be performed after refueling when

few noncondensible gases remain in the reactor coolant. The

temperature limitation of 212 deg.F will ensure that water, not

steam, would be emitted from the postulated leak or pipe break. In

addition, under these conditions, stored energy is sufficiently low

that even with loss of inventory following a recirculation line

break, core coverage would be maintained by the low pressure

emergency core cooling systems required per Specification 3.5.H and

the fuel would not exceed its peak clad temperature limit. As a

result, the potential for failed fuel and a subsequent increase in

reactor coolant activity is minimized and significant releases of

radioactive material to the environment would not be expected to

occur. Therefore, these changes will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

(2) The proposed changes do not involve a physical alteration of

the plant (no new or different type of equipment will be installed)

or changes in parameters governing normal operation and will not

alter the method used by any system to perform its design function.

The proposed changes to not allow plant operation in any mode that

is not already evaluated and will still ensure Secondary Containment

Integrity is maintained when required. Thus, these changes do not

create the possibility of a new or different kind of

[[Page 20860]]

accident from any accident previously evaluated.

(3) The proposed changes to Secondary Containment Integrity

requirements have no impact on any safety analysis assumptions.

Secondary Containment Integrity will be maintained as assumed in the

safety analyses and as stated in current Bases 3.7.B and 3.7.C. Not

requiring Secondary Containment Integrity when the reactor coolant

system is not vented in the Cold Shutdown condition or the Refuel

Mode does not involve significant reduction in a margin of safety

since no mechanism exists to impart additional fission products into

the reactor coolant. Under these conditions, activities for which

the reactor coolant system would not be vented would be strictly

controlled and monitored. As a result, leaks or pipe breaks would

typically be detected before significant inventory loss occurred.

These activities would typically be performed after refueling, at

low decay levels, and with reactor coolant temperature less than or

equal to 212 deg.F. In addition, under these conditions, stored

energy in the reactor core is very low. The reactor pressure vessel

would rapidly depressurize in the event of a large primary system

leak and the low pressure emergency core cooling systems required

per Specification 3.5.H under these conditions would be adequate to

keep the core flooded. This would ensure that the fuel would not be

uncovered and would not exceed the peak clad temperature limit.

As a result, the potential for failed fuel and a subsequent

increase in reactor coolant activity is minimized and significant

releases of radioactive material to the environment would not be

expected to occur. Therefore, these changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: R.K. Gad, III, Ropes and Gray, One

International Place, Boston, MA 02110-2624.

NRC Project Director: Susan Frant Shankman.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: April 4, 1996.

Description of amendment request: The proposed amendment would

revise the surveillance requirements for control rod over-travel to

remove the specific testing methodology from the Technical

Specifications to administratively controlled documents.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The control rod drive mechanism over-travel is not

considered to be the initiator of any previously analyzed accident.

Verification of coupling of the control rods and drive mechanisms is

performed by other means and continues to be required in the same

manner, so there is no significant increase in the probability of a

rod drop accident. The over-travel indication is also not considered

in the mitigation of consequences of any previously analyzed

accident, and the removal of a specific surveillance of the

indication will not affect the response of the control rods or the

reactor protection system to these accidents. Therefore, this change

will not significantly increase the probability or consequences of

any previously analyzed accident.

(2) The proposed change does not necessitate a physical

alteration of the plant (no new or different type of equipment will

be installed) nor changes in parameters governing normal plant

operation. The proposed change will continue to provide effective

methods to assure the control rods and their drive mechanisms are

coupled and preserve the safety functions associated with the

prevention or automatic mitigation of design basis accidents. Thus,

this change does not create the possibility of a new or different

kind of accident from any accident previously evaluated.

(3) The proposed changes continue to provide an appropriate

method for verification of the capability of the over-travel

indication to perform its function. Therefore, this change will not

significantly reduce a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: R.K. Gad, III, Ropes and Gray, One

International Place, Boston, MA 02110-2624.

NRC Project Director: Susan Frant Shankman.

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

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

Date of amendment request: April 15, 1996.

Description of amendment request: The proposed changes will clarify

the applicability of the quadrant power tilt ratio (QPTR) requirements.

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:

Operation of Surry Power Station in accordance with the proposed

Technical Specifications change will not:

1. Involve a significant increase in the probability of

occurrence or the consequences of an accident previously evaluated.

The application of the QPTR limits, as proposed, will assure

that the gross core radial power distribution remains consistent

with design limits above 50% power. At or below 50% rated thermal

power, there is insufficient stored energy in the fuel or

insufficient energy being transferred to the reactor coolant to

require implementation of a QPTR limit on the distribution of core

power. Therefore, the proposed change to clarify the applicability

of the QPTR requirements has no impact on the probability of an

accident occurrence and does not increase the consequences of any

design basis accident.

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

from any accident previously evaluated.

There are no plant modifications or changes in methods of plant

operation introduced by the proposed change. The change would limit

the application of QPTR limits to operation at power levels >50% to

preclude core power distributions from occurring which would violate

fuel design criteria previously analyzed. At or below 50% rated

thermal power, there is no impact to core power distributions which

could affect the fuel design criteria. Therefore, the proposed

change does not create the possibility for an accident or

malfunction of a different type than that previously evaluated in

the safety analysis report.

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

The proposed change only affects the applicability of the QPTR

limits. The QPTR limits remain unchanged to preclude any violation

of previously analyzed fuel design criteria. Adherence to the QPTR

limits, hot channel factors, and applicable Limiting Conditions for

Operation will continue. Therefore, the margin of safety as

described in the Bases Section of any part of the Technical

Specifications is not reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

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

Williams,

[[Page 20861]]

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

23219.

NRC Project Director: Eugene V. Imbro.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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

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

Date of amendment request: April 3, 1996.

Description of amendment request: The proposed amendments would

revise the hydrogen mitigation system Technical Specifications (TS).

The change would provide that, if neither the Train A or Train B

igniter is operable in any one containment region, then there is an

allowance of 7 days to restore one hydrogen igniter to OPERABLE status,

or be in Hot Shutdown within the next 6 hours. This would be consistent

with the guidance of the Standard TS for Westinghouse plants, NUREG-

0431.

Date of publication of individual notice in Federal Register: April

16, 1996 (61 FR 16649).

Expiration date of individual notice: May 16, 1996.

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

Black Street, Rock Hill, South Carolina.

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

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

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

Date of amendment request: January 22, 1996, as supplement by

letter dated April 4, 1996.

Brief description of amendments: The proposed amendment would

modify the steam generator tube plugging criteria in Technical

Specification 3/4.4.5, Steam Generators, and the allowable leakage in

Technical Specification 3/4.4.6.2, Operational Leakage, and the

associated Bases. The proposed amendment would allow the implementation

of steam generator voltage-based repair criteria for the tube support

plate (TSP)/tube intersections for Unit 1.

Date of individual notice in the Federal Register: April 16, 1996

(61 FR 16651)

Expiration date of individual notice: May 16, 1996.

Local Public Document Room location: Wharton County Junior College,

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

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: March 14, 1996.

Description of amendment request: The proposed amendment would

revise the Technical Specifications for Indian Point Nuclear Generating

Unit No. 3 to allow a one-time extension of the test intervals for the

pressurizer safety valve setpoint and snubber functional testing that

is due in May 1996.

Date of publication of individual notice in Federal Register: April

3, 1996 (61 FR 14835)

Expiration date of individual notice: May 3, 1996.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document rooms for

the particular facilities involved.

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

Plant, Middlesex County and Northeast Nuclear Energy Company, et al.,

Docket Nos. 50-245, 50-336, and 50-423, Millstone Nuclear Power

Station, Units 1, 2, and 3, New London County, Connecticut

Date of application for amendments: November 22, 1995.

Brief description of amendments: The amendments delete from the

Technical Specifications certain review responsibilities of the Plant

Operations Review Committee and the Site Operations Review Committee

relating to the Emergency Plan and the Security Plan and their

respective implementing procedures. The proposed changes are consistent

with the guidance of Generic Letter 93-07.

Date of issuance: April 24, 1996

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

within 60 days.

Amendment Nos.: 189, 94, 196, and 128

Facility Operating License Nos. DPR-61, DPR-21, DPR-65, AND NPF-49:

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 14, 1996 (61

FR 5812)

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 24, 1996.

No significant hazards consideration comments received: No.

[[Page 20862]]

Local Public Document Room location: Russell Library, 123 Broad

Street Middletown, Connecticut 06457, for the Haddam Neck Plant, and

the Learning Resources Center, Three Rivers Community-Technical

College, 574 New London Turnpike, Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, CT 06385

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