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 January 4, 1997, through January 16, 1997.

The last biweekly notice was published on January 15, 1997 (62 FR

2185).

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 Chief, 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 February 28, 1997, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

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

Building, 2120 L Street, NW., Washington, DC 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.

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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. 2

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.

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: December 30, 1996

Description of amendment request: The amendment revises (1)

chemistry data (nickel content) shown on Technical Specification (TS)

Figures 3.4-2 and 3.4-3 for TS 3/4.4.9, ``Pressure/Temperature

Limits,'' and (2) the associated Bases 3/4.4.9 to reflect changes to

chemistry and material properties and changes to comply with recent

U.S. Nuclear Regulatory Commission (NRC) rule changes to 10 CFR 50,

Appendix G.

Basis for proposed no significant hazards 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 change does not involve a significant hazards consideration

for the following reasons:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

There are no physical changes to any plant equipment created by

the proposed changes. The chemistry and material property changes do

not impact the ability of the reactor vessel to maintain [its]

pressure boundary integrity as previously evaluated. The decrease in

EOL USE [End-of-Life Upper Shelf Energy] for weld heat 5P6771 is

relatively minor and remains above the required value that has been

prescribed by the NRC to provide the necessary level of ductility

assumed for reactor vessel integrity evaluations. Therefore, the

accident initiating and mitigating aspects of the pressure vessel

are not affected. In addition, neither the proposed change requiring

the ISLH [In-Service Leak and Hydrostatic] test to be complete

before the core is critical nor the proposed change allowing fuel in

the reactor vessel during ISLH affects any accident initiating

mechanisms. The proposed change requiring the ISLH test to be

completed before the core is critical will not increase the

consequences of previously evaluated accidents because it

conservatively assures the core is subcritical. Although the

proposed change allows fuel in the vessel during ISLH utilizing the

ISLH Pressure-Temperature (P-T) limits, the consequences of a

pressure boundary leak have not changed because ISLH testing is

already allowed using the normal plant P-T limits. In addition, the

ISLH will be required to be completed before the core is allowed to

go critical. The consequences of a leak with fuel in the vessel

during ISLH are the same using either the normal P-T limits or the

ISLH limits.

Therefore, there would be no increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

There are no physical changes to any plant equipment or new

components created by the proposed changes. The chemistry and

material property changes do not impact the pressure boundary

integrity of the reactor vessel. The decrease in EOL USE for weld

heat 5P6771 is relatively minor and remains above the required value

that has been prescribed by the NRC to provide the necessary level

of ductility assumed for reactor vessel integrity evaluations.

Therefore, the accident initiating aspects of the pressure vessel

are not affected. In addition, neither the proposed change requiring

the ISLH test to be complete before the core is critical nor the

proposed change allowing fuel in the reactor vessel during ISLH

creates any new accident initiating mechanisms.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

[[Page 4343]]

The changes in chemical and material properties do not adversely

affect any reactor vessel integrity evaluations, such as PTS

[Pressurized Thermal Shock] or P-T limits. The USE for weld heat

5P6771 does decrease slightly as described in TS Bases Table B 3/

4.4-1. However, the predicted EOL USE remains above the value

prescribed in 10 CFR 50, Appendix G and is not a significant

reduction in the margin of safety. With regard to the proposed

changes allowing fuel in the reactor vessel during ISLH, the

existing TS Bases specifically state that fuel is not to be in the

reactor vessel when the ISLH P-T curve is utilized. However, this

change is consistent with the revised 10 CFR 50, Appendix G rule and

as such, is not a significant reduction in the margin of safety.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: Mark Reinhart, Acting

Commonwealth Edison Company, Docket No. 50-010, Dresden Nuclear

Generating Station, Unit 1, Grundy County, Illinois

Date of amendment request: October 23, 1996

Description of amendment request: The proposed change would amend

the Dresden Unit 1 Appendix A Technical Specifications (TS). The

proposed amendment is a complete revision of the TS to the same format

as Dresden Unit 2/3 TS.

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

1. Will operation of the facility according to this proposed

change involve a significant increase in the probability or

consequences of an accident previously evaluated?

No. In general the proposed amendment represents the conversion

of current requirements to a more generic format, or the addition of

requirements which are based on the current safety analysis

(Decommissioning Plan). Implementation of these changes will not

reduce reliability of equipment assumed to operate in the current

safety analysis (Decommissioning Plan), or will provide continued

assurance that specified parameters remain within their acceptance

limits, and as such, will not significantly increase the probability

or consequences of a previously evaluated accident.

Some of the proposed changes represent minor curtailments of the

current requirements which are based on generic guidance or

previously approved provisions for other stations. The proposed

amendment for Dresden Station Unit 1's Technical Specifications in

general is based on STS [Standard Technical Specifications]

guidelines or NRC accepted changes to other facilities such as

Trojan or San Onofre Unit 1. Any deviations from STS requirements do

not significantly increase the probability or consequences of any

previously evaluated accidents for Dresden Station Unit 1. The

proposed amendment is consistent with the current safety analysis

(Decommissioning Plan) and has been previously determined to

represent sufficient requirements for the assurance and reliability

of equipment assumed to operate in the safety analysis

(Decommissioning Plan), or provide continued assurance that

specified parameters remain within their acceptance limits. As such,

these changes will not significantly increase the probability or

consequences of a previously evaluated accident.

2. Will operation of the facility according to this proposed

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

from any accident previously evaluated.

No. In general, the proposed amendment represents the conversion

of current requirements to a more generic format, or the addition of

requirements which are based on the current safety analysis

(Decommissioning Plan). Others represent minor curtailments of the

current requirements which are based on generic guidance or

previously approved provisions for other stations. These changes do

not involve revisions to the design of the station. Some of the

changes may involve revision in the operation of the station;

however, these provide additional restrictions which are in

accordance with the current safety analysis (Decommissioning Plan).

The proposed amendment for Dresden Station Unit 1's Technical

Specifications in general is based on STS guidelines or NRC accepted

changes to other facilities such as Trojan or San Onofre Unit 1. The

proposed amendment has been reviewed for acceptability at the

Dresden Nuclear Power Station considering similarity of system or

component design versus the STS of later operating plants. Any

deviations from STS requirements do not create the possibility of a

new of different kind of accident previously evaluated for Dresden

Station, Unit 1. No new modes of operation are introduced by the

proposed changes. The proposed changes maintain at least the present

level of operability. Therefore, the proposed changes do not create

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

previously evaluated.

3. Will operation of the facility according to this proposed

change involve a significant reduction in a margin of safety?

No. In general, the proposed amendment represents the conversion

of current requirements to a more generic format, or the addition of

requirements which are based on the current safety analysis

(Decommissioning Plan). Others represent minor curtailments of the

current requirements which are based on generic guidance or

previously approved provisions for other stations. Some of the later

individual items may introduce minor reductions in the margin of

safety when compared to the current requirements. However, other

individual changes are the adoption of new requirements which will

provide significant enhancement of the reliability of human

performance assumed in the safety analysis (Decommissioning Plan),

or provide enhanced assurance that specified parameters remain

within their acceptance limits. These enhancements compensate for

the individual minor reductions, such that taken together, the

proposed changes will not significantly reduce the margin of safety.

The proposed amendment to Technical Specification Section 6.0

implements present requirements, or the intent of present

requirements in accordance with the guidelines set forth in the STS.

Any deviations from STS requirements do not significantly reduce the

margin of safety for Dresden Station. The proposed changes are

intended to improve readability, usability, and the understanding of

technical specification requirements while maintaining acceptable

levels of safe operation. The proposed changes have been evaluated

and found to be acceptable for use at Dresden based on system

design, safety analysis requirements and operational performance.

Since the proposed changes are based on NRC accepted provisions at

other operating plants that are applicable at Dresden and maintain

necessary levels of system or component reliability, the proposed

changes do not involve a significant reduction in the margin of

safety.

The NRC staff has reviewed the analysis of the licensee 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: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Seymour H. Weiss

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 2, 1996

Description of amendment request: The proposed amendments would

[[Page 4344]]

revise Technical Specification 3/4.4.2 to reduce the number of required

Safety/Relief Valves (SRVs). This change will support a modification to

remove five of the currently installed SRVs due to the current excess

capacity, and to reduce maintenance costs and worker radiation dose.

The current requirement for 17 of the 18 installed SRVs to be operable

would be changed to require 12 of the 13 installed SRVs to be operable.

Basis for proposed no significant hazards 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 probability of an accident previously evaluated will not

increase as a result of this change, because the change in valve

configuration, and the accompanying piping modification does not

alter any of the initiators of an accident or cause them to occur

more frequently. The piping modifications will be performed

consistent with the current piping classifications for the affected

components. Removal of the SRVs will not impact the ability of the

remaining SRVs to perform their functions, as described below.

The consequences of an ASME Overpressurization Event are not

significantly increased and do not exceed the previously accepted

licensing criteria for this event. General Electric (GE) has

calculated the revised peak vessel pressure for LaSalle Station to

be 1341 psig, which is below the 1375 psig criterion of the ASME

Code for upset conditions, referenced in Section 5.2.2,

Overpressurization Protection, of the Updated Final Safety Analysis

Report (UFSAR), and NUREG-0519 (Safety Evaluation Report related to

the operation of LaSalle County Station, Units 1 and 2, March 1981),

and Section 15.2-4, Closure of Main Steam Isolation Valves (BWR) of

NUREG-0800 (Standard Review Plan). The consequences of this event

will continue to be verified on a cycle-specific basis, beginning

with LaSalle Unit 1 Cycle 9 (L1C9). These analysis results will be

approved as part of the normal reload licensing 10 CFR 50.59

processes.

GE has also performed an analysis of the limiting Anticipated

Transient Without Scram (ATWS) event, which is the MSIV Closure Event

(MSIVC). This analysis calculated the peak vessel pressure to be 1378

psig, which is well below the 1500 psig criterion of the ASME Code for

emergency conditions. General Electric has verified that these results

will not be impacted with the introduction of Siemens fuel.

The conclusions given in the safety analyses with regards to

primary containment dynamic loads, main steam piping loads, Loss-of-

Coolant Accident (LOCA) impact, Minimum Critical Power Ratio (MCPR)

impact and SRV availability also show that current accident and

transient analyses are not impacted by this change beyond those

reanalyzed by GE and discussed above.

There is no increase in the amount or types of radioactive

release for any of the affected accidents or transients.

Therefore, there is not a significant increase in the

consequences of an accident previously evaluated.

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

from any accident previously evaluated because:

The as-left SRV piping configuration will continue to be

consistent with the current classifications for these piping and

supports, and have been evaluated by Sargent and Lundy analyses.

This ensures no different types of events may be caused by piping

failures at these locations. This is the only physical modification

proposed by this submittal, and it will not create the possibility

of a new or different kind of accident from those previously

evaluated. Other systems are not modified with this change and have

been shown in this submittal to continue to function as intended

with the new system configuration, with the exception of the

abandoned discharge line snubbers which may be replaced with struts,

except where they will be retained as snubbers due to thermal

expansion requirements. The changed supports are required to

function only as struts with the revised piping. Consideration and

evaluation of this function ensure no new or different accidents are

created.

3) Involve a significant reduction in the margin of safety

because:

While the calculated peak vessel pressures for the ASME

Overpressurization Event and the MSIVC ATWS Event are increased due

to the proposed SRV removals, the new peak pressures remain below

the respective licensing acceptance limits associated with these

events.

The actual cycle-specific reload analysis of the ASME

Overpressurization Event will be verified to be within the licensing

acceptance limit for that event prior to each cycle startup, as

required in the normal reload 10CFR50.59 process. These licensing

acceptance limits have been previously evaluated as providing a

sufficient margin of safety. For other accidents and transients,

including suppression pool loadings, the SRV removals have a

negligible, if any, effect on the results, so the margin of safety

is preserved.

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

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

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

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

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

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

Date of amendment request: August 14, 1996

Description of amendment request: The proposed amendment would

revise Technical Specification Sections 3.3 (Engineered Safety

Features) and 6.9.1.9 (Core Operating Limits Report (COLR)); the basis

of Section 3.3, 3.6 (Containment) and 3.10 (Control Rods). These

changes would incorporate the best estimate approach into the licensing

basis for the Indian Point Unit No. 2 large break loss-of-coolant

accident (LOCA) analysis.

Basis for proposed no significant hazards 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:

No physical changes are being made by this change. The plant

conditions assumed in the analysis are bounded by the design

conditions for all equipment in the plant. Therefore, there will be

no increase in the probability of a loss-of-coolant accident. The

consequences of a LOCA are not being increased. That is, it is shown

that the emergency core cooling system is designed so that its

calculated cooling performance conforms to the criteria contained in

50.46 paragraph b, that is it meets the five criteria listed in

Section II [see application dated August 14, 1996] of this

evaluation. No other accident is potentially affected by this

change. Therefore, neither the probability nor the consequences of

an accident previously analyzed is increased due to the proposed

change.

2) Does the proposed license amendment create the possibility of

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

Response:

There are no physical changes being made to the plant. No new

modes of plant operation are being introduced. The parameters

assumed in the analysis are within the design limits of existing

plant equipment. All plant systems will perform equally during the

response to a potential accident. Therefore, the possibility of a

new or different kind of accident than previously analyzed will not

be increased.

3. Does the proposed amendment involve a significant reduction

in the margin of safety?

Response:

It has been shown that the analytic technique used in the

analysis realistically describes the expected behavior of the Indian

Point Unit No. 2 reactor system during a

[[Page 4345]]

postulated loss of coolant accident. Uncertainties have been

accounted for as required by 10 CFR 50.46. A sufficient number of

loss of coolant accidents with different break sizes, different

locations and other variations in properties have been analyzed to

provide assurance that the most severe postulated loss of coolant

accidents were calculated. It has been shown by the analysis that

there is a high level of probability that all criteria contained in

10 CFR 50.46 paragraph b) are met. Therefore the proposed amendment

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

New York, New York 10003.

NRC Project Director: S. Singh Bajwa, Acting Director

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

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

Date of amendment request: August 21, 1996

Description of amendment request: The proposed amendment would

change the licensee's Technical Specifications (TSs) Section 3.3.G

(Hydrogen Recombiner System and Post-Accident Containment Venting

System), the basis for Section 3.3.G, and Section 4.4, Table 4.4-1

(Containment Isolation Valves). The change would remove the existing

flame-type hydrogen recombiners, its supporting equipment, and replace

it with passive autocatalytic recombiners (PARs). In addition, the

design basis analysis of post-accident hydrogen generation would be

recalculated.

Basis for proposed no significant hazards 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?

Neither the probability nor the consequences of a post-LOCA

[loss-of-coolant accident] combustible gas accident are increased by

the change in recombiners or in the change to hydrogen generation

analysis. The probability of a 10 CFR 59.44 type LOCA is not

affected. The consequences of such an accident are not significantly

changed.

Accidents associated with failure of the flame-recombiner flue

(hydrogen/oxygen) system as well as with failure of the flame-

recombiner containment isolation valves have been eliminated.

No other accident is potentially affected by this change.

2) Does the proposed license amendment create the possibility of

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

No new modes of plant operation are being introduced other than

elimination of operation of the flame-type recombiners and

associated support equipment. Recombiner failure is believed to be

far less likely with the PAR design but in any event, the

containment vent system is being maintained in its current role as

backup to recombiner systems. All other plant systems will perform

equally during the response to a potential accident. Therefore, the

possibility of a new or different kind of accident than previously

analyzed will not be increased.

3) Does the proposed amendment involve a significant reduction

in the margin of safety?

The proposed amendment involves margin in the hydrogen

flammability limit, in the hydrogen generation assumptions and in

the number of PAR devices assumed. Furthermore, sensitivity analysis

on PAR effectiveness indicates that additional margin exists for

success even with degraded PAR performance. It has been shown by the

analysis that the criteria of 10 CFR 50.44(d) can be met with

margin. Therefore, the proposed amendment 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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

New York, New York 10003.

NRC Project Director: S. Singh Bajwa, Acting Director

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

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

Date of amendment request: August 22, 1996

Description of amendment request: The proposed amendment would

revise the licensee's Technical Specification Sections 3.3 and 4.5

(Engineered Safety Features). The proposed revision would delete the

requirement to utilize sodium hydroxide (NaOH) as an additive in the

posted-accident containment spray system.

Basis for proposed no significant hazards 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:

...consistent with the Commission's criteria in 10 CFR 50.92, we

have determined that the proposed change does not involve a

significant hazards consideration because the operation of Indian

Point Unit No. 2 in accordance with this change would not:

1) involve a significant increase in the probability or

consequences of an accident previously evaluated. The proposed

revisions are based on conservative analyses utilizing new, approved

methodologies. The analysis shows the sodium hydroxide spray

additive can be removed without significantly affecting the

radiological consequences of a postulated LOCA [loss-of-coolant

accident] and that the calculated off-site doses would remain within

the 10 CFR 100 guidelines. In order to maintain acceptable pH levels

in the recirculating ECC [emergency core cooling] solution, baskets

of trisodium phosphate will be stored in strategic locations in

containment.

2) create the probability of a new or different kind of accident

from any accident previously evaluated. The proposed change allows

the containment safeguards to mitigate the consequences of a design

basis LOCA in a manner equivalent to that previously approved.

3) involve a significant reduction in a margin of safety. With

the proposed change, all safety criteria previously evaluated are

still met and remain conservative.

Therefore, based on the above, we conclude that the proposed

changes do not constitute a significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

New York, New York 10003.

NRC Project Director: S. Singh Bajwa, Acting Director

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

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

Date of amendment request: January 3, 1997

Description of amendment request: The proposed amendments would

eliminate from various parts of the Technical Specifications any

requirement for the low steam pressure

[[Page 4346]]

signal as an initiator of safety injection. The licensee stated that

the function of the signal is adequately performed by other signals

(such as the low pressurizer pressure signal).

Basis for proposed no significant hazards 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 will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change, to delete the SI [safety injection] signal

on low steam line pressure, will only prevent an unnecessary SI

actuation as an event occurs which involves secondary system

depressurization. No consequences will significantly increase,

because for each event previously analyzed it has been shown that

either SI on low steam pressure is not demanded, or that another SI

signal (e.g., low pressurizer pressure) is generated in sufficient

time to meet applicable acceptance criteria. The probability of an

accident will not increase.

2. The proposed change will not create the possibility of any

new accident not previously evaluated.

The initiation of SI on a low steam line pressure signal may

occur during events which involve a depressurization of the

secondary side, including excessive auxiliary feedwater addition.

There are other SI initiation signals which will accomplish this

same function if needed. Removing this actuation signal will not

create any new failure modes or necessitate any new hardware

configurations (other than the deletion of the signal itself). No

new accident scenarios are created.

3. There is no significant reduction in a margin of safety.

Analysis has shown that for any transient for which SI would

have occurred on low steam line pressure, transient response is

maintained within acceptable limits. Steam line break mass and

energy releases inside containment do not violate the existing

environmental qualification envelope. Steam line breaks outside

containment are not adversely affected by this 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

proposed amendments involve no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,

Unit Nos. 1 (ANO-1), Pope County, Arkansas

Date of amendment request: November 26, 1996

Description of amendment request: Change Reactor Coolant System

Pressure and Temperature Curvers

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

Criterion 1 - Does not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The proposed change revises the pressure/temperature limits in

accordance with the 10 CFR 50.60 requirements or in accordance with

Code Case N-514. This approach utilizes the latest NRC guidelines

relative to estimating neutron irradiation damage of the reactor

vessel, as well as maintaining conservative limits with respect to

the low temperature overpressure protection (LTOP) system.

Therefore, this change does not involve a significant increase in

the probability or consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

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

different kind of accident from any previously evaluated since it

does not introduce new systems, failure modes or plant

perturbations. Therefore, this change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in Margin

of Safety.

The proposed change will not involve a significant reduction in

the margin of safety since the proposed pressure/temperature

limitations have been developed consistent with the requirements of

10 CFR 50.60. The operational limits have been developed to maintain

the necessary margins of safety through 32 effective full power

years using methodologies previously reviewed and approved by the

NRC. The objective of these limits is to prevent non-ductile failure

during any normal operating condition, including anticipated

operational occurrences and system hydrostatic tests.

The LTOP safety factors are based on reanalyzed conditions for

32 effective full power years 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% of the pressure determined to satisfy Appendix G,

paragraph G-2215 of ASME Section XI, Division 1 as a design limit,

instead of 100% required by Section XI. This proposed alternative is

acceptable because the Code Case recognizes the conservatism of the

ASME Appendix G curves and allows establishing a LTOP setpoint which

retains an acceptable margin of safety while maintaining operational

margins for reactor coolant pump operation at low temperatures and

pressures. The Code Case provides an acceptable margin of safety

against flaw initiation and reactor vessel failure, and reduces the

potential for an undesired LTOP actuation. The application of Code

Case N-514 for ANO-1 will ensure an acceptable level of safety.

Therefore, this change does not involve a significant reduction in

the margin of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: October 7, 1996

Description of amendment request: Modify Plant Protection System

Test Interval to 123 days.

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

Criterion 1 - Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The proposed changes included in this amendment request are

being made to surveillance intervals, allowances to use CISAM

elements and various administrative changes. These changes do not

alter the functional characteristics of any plant component and do

not allow any new modes of operation of any components. These

changes do not involve a significant increase in the probability of

any event initiator to occur. Therefore, this amendment request does

not involve a significant increase in the probability of any

accident previously evaluated.

Increasing the surveillance interval for the RPS and ESFAS

instrumentation has two principal effects with opposing impacts on

risk. The first impact is a slight increase in core damage frequency

that results from the increased unavailability of the

[[Page 4347]]

instrumentation in question from the extended testing interval. The

unavailability of the tested instrumentation components is

translated to result in a failure of the reactor to trip, an

anticipated transient without a scram, or a failure of the

appropriate engineered safety feature to actuate when required. The

opposing impact on risk is the corresponding reduction in core

damage frequency that would result due to the reduced exposure of

the plant to test induced transients.

Representative fault tree models were developed for ANO-2 and

the corresponding core damage frequency increases and decreases were

quantified in CEN-327 and CEN-327 Supplement 1. The NRC staff found

that changes in the RPS unavailabilities that result from extending

the surveillance test interval (STI) from 30 days to 90 days were

not considered to be significant. Estimates of the reduction in

scram frequency from the reduction in test induced scrams and the

corresponding reduction in core damage frequency were found

acceptable. Sequential testing intervals of 90 days were found to

result in a net reduction in risk.

CE NPSD-576 employed the same methodology used in CEN-327 and

its supplement to evaluate the impact of extending the surveillance

intervals from monthly sequential testing to every four months

(triannual) on a staggered test basis. The corresponding changes in

RPS and ESFAS unavailabilities are quantified in CE NPSD-576 and are

shown to be less than their counterparts in CEN-327 and its

supplement. Thus, triannual staggered testing should be acceptable

as it results in lower RPS and ESFAS unavailabilities than for a 90

day test interval with sequential testing which has been found to be

acceptable to the NRC.

The TS amendment request provided the option to use cycle

independent shape annealing matrix (CISAM) elements. The CISAM

elements will be validated during startup testing and will be

required to meet additional acceptance criteria as well as that used

for the cycle specific shape annealing matrix (SAM) elements. If the

CISAM is determined to be no longer valid, a cycle specific SAM will

be calculated and used in the CPCs. Therefore, the CPCs will operate

as designed and this change will not affect the consequences of any

accident previously evaluated.

The CPC addressable constant surveillance requirements and the

various administrative changes affected by this TS change do not

affect the consequences of any accident previously evaluated.

Therefore, these changes do not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

This amendment request does not involve any changes in equipment

and will not alter the manner in which the plant will be operated.

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

or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

The RPS/ESFAS extended testing interval yields no significant

reduction in the margin to safety. The instrument drift occurring

over the proposed STI will not cause the setpoint values to exceed

those assumed in the safety analysis and specified in the TS. There

are no changes to equipment or plant operations that will result

from this change. The implementation of these proposed changes is

expected to result in an overall improvement in safety due to the

fact that reduced testing will result in fewer inadvertent trips,

less frequent actuation of EFAS components, and less frequent

distraction of the operations personnel.

The CPC addressable constant surveillance interval extension

included in this amendment request is consistent with the

methodology found in NUREG-1432, ``Standard Technical Specifications

Combustion Engineering Plants'' (ISTS). Requiring the addressable

constant verification to be performed as part of the CPC channel

functional test should detect an error in these constants prior to

restoring the channel to operable status instead of allowing the

error to go undetected until the next surveillance period. Although

the surveillance interval is extended by this TS change, this change

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

The CPC CISAM elements and the various administrative changes

included in this TS change do not involve a significant reduction in

the margin of safety.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: December 19, 1996

Description of amendment request: Change Request Concerning

Addition to the Core Operating Limit Report References

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

Criterion 1 - Does not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The proposed change to add the technical manual for the

Combustion Engineering Nuclear Transient Simulation (CENTS) code to

the Core Operating Limits Report (COLR) references is administrative

in nature. The CENTS code has been reviewed and approved by the NRC.

The physical design or operation of the plant is not impacted by

this proposed change. The proposed change does not adversely impact

transient analysis assumptions or results. The COLR-related safety

analyses will continue to be performed utilizing NRC-approved

methodologies, and specific reload changes will be evaluated under

the provisions of 10CFR50.59. Therefore, this change does not

involve a significant increase in the probability or consequences of

any accident previously evaluated.

Criterion 2 - Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

The proposed change to reference the NRC-approved CENTS code is

administrative in nature. No physical alterations of plant

configuration, changes to plant operating procedures, or operating

parameters are proposed. No new equipment is being introduced, and

no equipment is being operated in a manner inconsistent with its

design. The COLR-related safety analyses will continue to be

performed utilizing NRC-approved methodologies. A 10CFR50.59 safety

review will continue to be performed to evaluate specific reload

changes. Therefore, this change does not create the possibility of a

new or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

The proposed change to reference the CENTS code is

administrative in nature. Existing technical specification

operability and surveillance requirements are not reduced by the

proposed change. The cycle- specific COLR limits for future reloads

will continue to be developed based on NRC-approved methodologies.

Technical specifications will continue to require that the core be

operated within these limits and specify appropriate actions to be

taken if the limits are violated. The COLR-related safety analyses

will continue to be performed utilizing NRC-approved methodologies,

and specific reload changes will be evaluated per 10CFR50.59.

Therefore, this change does not involve a significant reduction in

the margin of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve a significant

hazards consideration.

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

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

satisfied. Therefore, the NRC staff

[[Page 4348]]

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: December 19, 1996

Description of amendment request: Change Request Concerning Power

Calibration Requirements

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

Criterion 1 - Does not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The proposed change will redefine the tolerance band allowed for

linear power level, the Core Protection Calculator (CPC) delta T

Power, and CPC nuclear power signals. Changing the tolerance range

from [plus or minus] 2% to between -0.5% and 10% between 15% and 80%

rated thermal power, will require more conservative tolerances than

are currently allowed. This change will ensure that the power

indications are more conservative relative to the existing safety

analyses. Therefore, this change does not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

Criterion 2 - Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

The proposed change to Technical Specification power calibration

tolerance limits are conservative relative to the current

requirements. This amendment request does not change the design or

operation of any plant systems or components. Therefore, this change

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

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in Margin

of Safety.

The allowed tolerance band for the linear power level, CPC delta

T power, and CPC nuclear power signals between 15 and 80% power has

been redefined. The new requirements are more conservative than the

tolerances that currently exist in the Technical Specifications.

This change will ensure that the power indications are more

conservative relative to the existing safety analyses. Therefore,

this change does not involve a significant reduction in the margin

of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: December 19, 1996

Description of amendment request: Change Request Concerning Reactor

Coolant System Volume

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

Criterion 1 - Does not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

This proposed change allows the relocation of the reactor

coolant system volume in the design features section of technical

specifications to the safety analysis report. Future changes will be

controlled under 10CFR50.59. This change is considered

administrative in nature. Appropriate values of reactor coolant

system volume are used in the safety analyses. This change does not

affect any system or component functional requirements. The

operation of the plant is not affected by this change.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

The relocation of existing requirements from the technical

specifications to another licensee controlled document is

administrative in nature. This change does not modify or remove any

plant design requirement. The proposed change will not affect any

plant system or structure, nor will it affect any system functional

or operability requirements. Therefore, no new failure modes are

introduced as a result of this change.

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

or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

The proposed amendment request relocates the coolant system

volume located in the technical specifications design feature

section to another licensee controlled document, the ANO-2 Safety

Analysis Report, which is controlled under 10CFR50.59. The proposed

change is administrative in nature because the design requirements

for the facility remain the same. The proposed change does not

represent a change in the configuration or operation of the plant.

Therefore, this change does not involve a significant reduction

in the margin of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve a significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Nos. 50-313 and 50-368, Arkansas

Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

Date of amendment request: December 19, 1996

Description of amendment request: Change Control Room Ventillation

System Requirements

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

Criterion 1 - Does not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The control room emergency ventilation and air conditioning

systems are not initiators of an accident previously evaluated.

Extension of the allowable outage time for one inoperable control

room emergency air conditioning system from 7 days to 30 days is

acceptable based on the low probability of an event occurring that

would require control room isolation and a concurrent or subsequent

failure of the remaining operable control room emergency air

conditioning system. An evaluation using probabilistic safety

assessment techniques has shown the frequency of this event to be an

acceptably low level (4.67E-6/yr). The ANO-1 surveillance

requirements for the control room emergency ventilation and air

[[Page 4349]]

conditioning system have been updated for consistency with the ANO-2

requirements and are consistent with RG 1.52, March 1978, Revision 2

and ASTM D3803-1989. The change in the ANO-2 Mode of Applicability

for the control room radiation monitoring instrumentation is

acceptable because the only identified accident scenario requiring

control room isolation on high radiation while in Modes 5 and 6 is

the fuel handling accident and this analysis shows that the dose

consequences to the control room operators are acceptable in the

event of a fuel handling accident, assuming that the normal control

room ventilation system is properly isolated. The remainder of the

changes have been made for consistency between the ANO-1 and ANO-2

TS and are considered to be more restrictive or administrative in

nature.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

The control room emergency ventilation and air conditioning

systems are not accident initiators. The proposed changes introduce

no new mode of plant operation and no new possibility for an

accident is introduced by modifying the ANO-1 surveillance testing

requirements for the control room emergency ventilation and air

conditioning systems.

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

or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

With the exception of the AOT extension and the relaxation of

the ANO-2 Mode of Applicability for the control room radiation

monitoring instrumentation, all the ANO-1 and ANO-2 changes are

considered administrative or more restrictive and are intended to

clarify and make consistent the requirements of the control room

emergency habitability equipment. Although the AOT extension does

involve an incremental reduction in the margin of safety due to

slight increase in the frequency of an event requiring control room

isolation, followed by failure of the operable emergency control

room chiller, a probabilistic safety assessment has shown this

slight increase in frequency (approximately 3.58E-6/yr) to be

acceptably low. The change in the ANO-2 Mode of Applicability for

the control room radiation monitoring instrumentation is acceptable

because the only identified accident scenario requiring control room

isolation on high radiation while in Modes 5 and 6 is the fuel

handling accident and this analysis shows that the dose consequences

to the control room operators are acceptable in the event of a fuel

handling accident, assuming that the normal control room ventilation

system is properly isolated.

Therefore, this change does not involve a significant reduction

in the margin of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket No.

50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County,

Georgia

Date of amendment request: December 3, 1996

Description of amendment request: The two proposed changes would

revise Technical Specification (TS) 2.1.1.2 for Hatch Nuclear Plant,

Unit 2, Safety Limit Minimum Critical Power Ratio (SLMCPR) values. The

revision is based upon unique plant evaluations for the current Cycle

13 and the use of General Electric (GE) GE-13 fuel, a 9 x 9 fuel

design, in the next Cycle 14. The proposed SLMCPRs for Hatch Unit 2 are

1.08 and 1.09 (single-loop operation) for the current Cycle 13, and

1.12 and 1.14 (single-loop operation) for Cycle 14.

The new SLMCPRs were calculated using NRC-approved methods and

interim implementing procedures. The SLMCPRs are set high enough to

ensure that greater than 99.9% of all fuel rods in the core avoid

transition boiling if the limit is not violated. The SLMCPRs

incorporate a margin for uncertainty in the core operating state for

uncertainties that are fuel-type dependent, including fuel bundle

nuclear characteristics, critical power correlation, and manufacturing

tolerances. These interim procedures were revised to incorporate the

following cycle-specific parameters: (1) Actual core loading, (2)

Conservative variations of projected control blade patterns, (3) Actual

bundle parameters (e.g., local peaking), and (4) Full cycle exposure

range.

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration which is presented

below:

1. Does the change involve a significant increase in the

probability of consequences of an accident previously evaluated?

The derivation of the revised SLMCPRs for Plant Hatch Unit 2 for

incorporation into the Technical Specifications, and its use to

determine cycle-specific thermal limits, were performed using NRC-

approved methods. Additionally, interim implementing procedures

incorporating cycle-specific parameters were used. Based upon the

use of these calculations, revised SLMCPRs cannot increase the

probability or severity of an accident. The basis of the SLMCPR

calculation is to ensure that 99.9% of all fuel rods in

the core avoid transition boiling if the limit is not violated. The

new SLMCPRs preserve the existing margin to transition boiling and

fuel damage in the event of a postulated accident. Thus, it can be

concluded that the probability of fuel damage is not increased and

the proposed Technical Specifications changes do not involve an

increase in the probability or consequences of an accident

evaluation.

2. Do the proposed changes create the possibility of a new or

different type of accident from any previously evaluated?

The SLMCPR is a Technical Specifications numerical value

designed to ensure that fuel damage from transition boiling does not

occur as a result of the limiting postulated accident. The SLMCPRs

were calculated using NRC-approved methods. Additionally, interim

procedures incorporating cycle-specific parameters were used in the

analysis. Therefore, the proposed changes do not create the

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

previously evaluated.

3. Do the proposed changes involve a significant reduction in

the margin of safety?

The margin of safety as defined in the Bases will remain the

same. The new SLMCPRs were calculated using NRC-approved methods

which are in accordance with the current fuel design and licensing

criteria. Additionally, interim implementing procedures, which

incorporate cycle-specific parameters were used. The SLMCPR remains

high enough to ensure that 99.9% of all fuel rods in the

core will avoid transition boiling if the limit is not violated,

thereby preserving the fuel cladding integrity. Therefore, the

proposed changes do not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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

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

NRC Project Director: Herbert N. Berkow

[[Page 4350]]

Georgia Power Company, Oglethorpe Power Corporation, Municipal

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

50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,

Appling County, Georgia

Date of amendment request: January 7, 1997

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) for Plant Hatch, Units 1 and

2, associated with Surveillance Requirement (SR) testing that requires

manually actuating every safety/relief valve (S/RV) during each unit

startup from a refueling outage. The proposed changes would provide an

alternate method of testing the S/RVs during shutdown conditions rather

than during unit startup as is currently done. This approach would

reduce valve leakage, thereby reducing the possibility of inadvertent

valve actuation and resultant plant transients. Additionally, deletion

of testing for the safety mode of the S/RVs is proposed since other

testing provides operability verification.

Furthermore, the licensee proposes relief from the applicable

requirements of the ASME OM Code (1995), Appendix I, paragraph I

3.4.1(d), which also requires manual actuating of S/RVs during unit

startup.

Current Unit 1 and Unit 2 SRs 3.5.1.12 and 3.6.1.6.1 require that

each S/RV be manually actuated at pressure conditions. Georgia Power

Company (GPC) proposes to revise SRs 3.5.1.12 and 3.6.1.6.1 that would

require the S/RVs to be manually actuated in the relief mode during a

plant outage before steam is generated. The solenoid valve would be

energized, the actuator would stroke, and the pilot rod lift would be

measured. This in-situ test would verify that, given a signal to the

solenoid, the pilot disc rod would lift. If steam were present, the

pilot disc would open and initiate opening of the main stage.

The licensee also proposes to delete current Units 1 and 2 SR

3.4.3.2, which also requires that each S/RV be manually actuated

because this test is not necessary to assure S/RV operability in the

safety mode since other tests, taken together, confirm the entire S/RV

assembly functions adequately.

Basis for proposed no significant hazards 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:

Georgia Power Company [GPC] has reviewed the proposed license

amendment request and determined its adoption does not involve a

significant hazards consideration. In support of this determination,

an evaluation of each of the three 10 CFR 50.92 standards follows.

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

the probability or consequences of an accident previously evaluated.

Since the proposed Technical Specifications changes and ASME

Code relief do not impose any physical changes to the S/RVs, their

design function is unaffected. The submittal only proposes changes

to the manner in which the S/RVs are tested. As discussed in

Enclosure 1 [of the licensee's submittal], the combination of

current S/RV testing and the proposed alternate

testing will continue to adequately demonstrate the operability

of the S/RVs for both the safety and relief modes. Under the

proposed testing requirements, it is expected that S/RV leakage will

decrease; thus, the probability of occurrence of an inadvertent S/RV

actuation is actually reduced.

FSAR [Final Safety Analysis Report] analyzed events, such as

MSIV [main steam isolation valve] closure, generator load reject,

turbine trip with failure of switchyard breakers to open, and

pressure regulator failure, take credit for the S/RVs mitigating the

consequences of these events. These proposed changes will not

increase the consequences of these events, since a series of S/RV

tests (on the bench and installed) will ensure all S/RV components

necessary to ensure valve opening will function. The S/RVs will

therefore be capable of performing their design functions.

Furthermore, reducing the number of manual actuations of the S/

RVs decreases the likelihood of a stuck open S/RV, which is an

analyzed event in the Hatch FSAR.

Therefore, the probability of occurrence and the consequences of

previously analyzed events are not increased.

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

or different kind of accident from any accident] previously

evaluated.

The proposed changes affect the manner in which S/RV operability

is verified in that one Technical Specifications SR [surveillance

requirement] is being deleted and two are being revised; however,

they do not affect the way the S/RVs are operated. The S/RVs will

not be operated or tested in a manner contrary to their design. As a

result, no new mode of operation is introduced. That is, the

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

kind of accident from any previously evaluated.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The present method of S/RV testing unnecessarily challenges the

valves, and is linked to S/RV degradation through pilot valve and/or

main valve leakage. This Technical Specifications change should

decrease S/RV leakage and improve S/RV reliability by reducing the

potential for spurious valve actuation at full power. In this sense,

the margin of safety is actually increased; e.g., the likelihood for

spurious S/RV actuation is reduced.

Deleting the test of installed S/RVs at rated temperature and

pressure will not significantly reduce the margin of safety for

events in which S/RV actuation is assumed, since each S/RV receives

a series of tests which insure each component necessary for

successful opening of the S/RV functions properly. Thus, the S/RV is

assured of opening in either the safety or the relief mode. For

example, at Wyle Labs, the valves undergo testing at operating steam

pressure. This test ensures operability of the pilot and main discs

and also verifies set pressure, reseat pressure, and main steam

stroke time. As noted previously, upon successful completion of

these tests, including verification of zero seat leakage, the valves

receive a written certification from the lab and are returned to

Plant Hatch for installation.

GPC further proposes that, upon installation, but before steam

is generated, the valves receive a test requiring the solenoid to be

energized. This test provides additional verification that the pilot

disc opens. The remaining segments of the S/RV tests verify the

ability of ADS and LLS logic to energize the solenoid.

In summary, this amendment does not involve a significant

reduction in the margin of safety, because of the reduction in S/RV

degradation, and because remaining tests confirm the valves will

function properly when required.

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: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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

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

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: December 16, 1996

Description of amendment request: The amendment request, if

approved, would reflect the change in the legal name of the operator of

TMI-1 from GPU Nuclear Corporation to GPU Nuclear Inc. and reflect in

the TMI-1 license and the Technical Specifications the registered trade

name of GPU Energy.

Basis for proposed no significant hazards 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:

GPU Nuclear Inc. has determined that the proposed TMI-1 license

amendment and technical specification change request

[[Page 4351]]

involve no significant hazards consideration as defined in 10 CFR

50.92 because:

Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The proposed amendment adds to the license and

the technical specifications the trade name of the Owners of TMI-1.

The change in the legal name of the operator of TMI-1 is a cosmetic

change made to reflect the name changes made throughout the GPU

family of companies. The name change has no impact on plant design

or operation.

Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated because no

new failure modes are created by the proposed changes. The use of a

common trade name for the Owners of TMI-1 and the change in the

legal name of the operator of TMI-1 has no impact on plant design or

operation. Thus, there is no creation of the possibility of a new or

different kind of accident from those previously evaluated.

Operation of the facility in accordance with the proposed

amendment will not involve a signficant reduction in a margin of

safety. The proposed amendment does not change any operating limits

for reactor operation.

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: Law/Government Publications

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

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Acting Project Director: Patrick D. Milano

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

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

Date of amendment request: August 4, 1995 as supplemented December

20, 1996 [AEP:NRC:1129E and 1129M]

Description of amendment request: The proposed amendment would

modify the technical specifications (T/S) to allow for repair of hybrid

expansion joint (HEJ) sleeves under redefined repair boundary limits.

This alternate plugging criterion would assess the integrity of parent

tube indications based on the degraded joint geometry, with reference

to the specific location of the flaw. The continued operability of the

HEJ sleeved tube would be based on the measured diameter difference, or

diameter delta (delta D), between the sleeve peak hardroll diameter and

the diameter of the sleeve adjacent to the parent tube flaw in the

upper joint.

Basis for proposed no significant hazards 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:

Conformance of the proposed amendments to the standards for a

determination of no significant hazard as defined in 10 CFR 50.92

(three factor test) is shown in the following.

(1) Operation of Cook Nuclear Plant unit 1 in accordance with

the proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The HEJ sleeved tube structural integrity limits defined by this

amendment provide for structural integrity consistent with the

guidance of RG 1.121. Tube structural integrity consistent with the

most limiting RG 1.121 loading is inherently provided by a measured

[delta D] of less than 1 mil, although the criterion specifies a

minimum of 3 mils must be verified. The structural integrity

characteristics of a postulated degraded parent tube with a 3 mil

[delta D] provides for axial restraint capability of more than

double the most limiting RG 1.121 loading, which indicates that the

postulated separated tube would not become axially displaced

relative to the sleeve during any plant condition.

Based on tube pull data from Cook Nuclear Plant and other plants

it is expected that TSP intersections would provide a substantial

axial restraint capability. This interaction is neglected in the

analysis of the criterion, and provides for extra safety margin.

Based on the destructive examination results for sections of HEJ

sleeved tubes removed in 1994 from another plant, the parent tube

flaw morphology is described as circumferentially oriented with

multiple initiation sites. This segmented morphology indicates that

the previously performed structural capability testing is

conservative. Additional axial load bearing capability is provided

by the segmented morphology since end cap loading would be

transmitted through the tube by the non-degraded ligaments of the

segmented crack network, and tube separation therefore, is not

likely or credible.

The consequences of any postulated failure of a sleeved tube to

which the criteria has been applied would be bounded by the current

steam generator tube rupture event discussed in the Cook Nuclear

Plant Final Safety Analysis Report (FSAR). Axial displacement of any

tube, sleeved or unsleeved, is bounded by approximately 1.1 inch. A

tube which experiences axial displacement by this amount would be

expected to exhibit a release rate well below the normal makeup

capacity. In order for a HEJ sleeved tube to exhibit reactor coolant

system release rates approaching the release rates assumed in the

FSAR the tube must be displaced by approximately 3 inches. In order

for the postulated separated tube to experience axial displacement

of any magnitude, it must be assumed that the HEJ hardroll provides

no structural benefit and that the tube-to-TSP interaction is

frictionless.

Postulated primary to secondary leakage during a main steam line

break event will be assessed against the limit of 8.4 gpm in the

faulted loop, calculated as part of the voltage based plugging limit

for tube support plate intersections. The total of all leakage

sources must be shown to be less than this value.

Application of the 3 mil [delta D] criterion (excluding eddy

current uncertainty) does not change existing reactor coolant system

flow conditions, therefore, existing LOCA analysis results will be

unaffected. Plant response to design basis accidents for the current

tube plugging and flow conditions are not affected by the repair

process; no new tube diameter restriction is introduced.

(2) The proposed license amendment does not create the

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

previously evaluated.

Application of the proposed 3 mil [delta D] HEJ sleeved tube

structural integrity criterion will not introduce significant or

adverse changes to the plant design basis. The 3 mil [delta D]

criteria provides for structural integrity of the HEJ sleeved tube

assembly which significantly exceeds the limiting RG 1.121 loading

condition. Under these conditions neither a single nor a multiple

tube rupture event is considered credible.

The general outline of the HEJ sleeve is unaffected, and the

application of the proposed criterion does not change the sleeve

configuration or size/shape. The application of the criterion also

does not represent a potential to affect other plant components.

(3) The proposed license amendment does not involve a

significant reduction in a margin of safety.

The proposed criterion has been shown to provide structural

integrity of the tube bundle consistent with the most limiting RG

1.121 tube integrity recommendations. In order for tube rupture to

occur, the degraded parent tube must experience a complete

circumferential separation and be subsequently axially displaced by

approximately 3 inches. The inherent structural integrity provided

by the interference fit of the HEJ in addition to the axial

restraint provided by tube support plate intersections above the HEJ

provides for structural integrity far exceeding the RG 1.121 loading

of 2264 lb. Even in the event that a degraded HEJ sleeved parent

tube were to experience axial displacement, the maximum amount of

displacement the tube could experience is bounded by 1.11 inch.

Postulating that the tube were to become displaced by this amount,

primary to secondary leakage would be limited to well less than the

normal makeup capacity due to the proximity between the

hydraulically expanded sleeve OD and tube ID.

Pulled HEJ sleeved tube samples from another plant with HEJ

sleeved tubes indicate that the crack morphology is described as

circumferentially oriented cracking with multiple initiation sites.

This segmented

[[Page 4352]]

morphology provides for additional structural margin not modeled in

the testing program.

Existing flow equivalency calculations for the HEJ sleeved tubes

will be unaffected by the application of the criterion.

Based on the preceding analysis it is concluded that operation

of Cook Nuclear Plant unit 1 following the application of the 3 mil

[delta D] HEJ sleeved tube structural integrity limit does not

increase the probability of an accident previously evaluated, create

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

accident previously evaluated, or reduce any margins to plant

safety. Therefore, the license amendment does not involve a

significant hazards consideration as defined in 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: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Gail H. Marcus

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit

No. 3, York County, Pennsylvania

Date of application for amendment: October 30, 1996

Description of amendment request: These amendments revise the

safety limit minimum critical power ratios (SLMCPRs) at Peach Bottom

Atomic Power Station, Unit 3.

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

1) The proposed TS [technical specification] changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The derivation of the cycle-specific SLMCPRs for incorporation

into the TS, and its use to determine cycle-specific thermal limits,

have been performed using USNRC [U.S. Nuclear Regulatory

Commission]-approved methods as discussed in ``General Electric

Standard Application for Reactor Fuel,'' NEDE-24011-P-A-11, and U.S.

Supplement, NEDE-24011-P-A-11-US, November 17, 1995 and interim

(reconfirmation) implementing procedures. This change in SLMCPRs

cannot increase the probability or severity of an accident.

The basis of the SLMCPR calculation is to ensure that greater

than 99.9% of all fuel rods in the core avoid transition boiling if

the limit is not violated. The new SLMCPRs preserve the existing

margin to transition boiling and fuel damage in the event of a

postulated accident. The fuel licensing acceptance criteria for the

SLMCPR calculation apply to PBAPS [Peach Bottom Atomic Power

Station], Unit 3, Cycle 11 in the same manner as they have applied

previously. The probability of fuel damage is not increased.

Therefore, the proposed TS changes do not involve an increase in the

probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The SLMCPR is a TS numerical value, designed to ensure that

transition boiling does not occur in 99.9% of all fuel rods in the

core during the limiting postulated accident. It cannot create the

possibility of any new type of accident. The new SLMCPRs are

calculated using USNRC-approved methods (General Electric

Standard Application for Reactor Fuel,'' NEDE-24011-P-A-11, and U.S.

Supplement, NEDE-24011-P-A-11-US, November 17, 1995) and interim

(reconfirmation) implementing procedures.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3) The proposed TS changes do not involve a significant

reduction in a margin of safety.

The margin of safety as defined in the TS Bases will remain the

same. The new SLMCPRs are calculated using USNRC-approved methods

(General Electric Standard Application for Reactor

Fuel,'' NEDE-24011-P-A-11, and U.S. Supplement, NEDE-24011-P-A-11-

US, November 17, 1995) and interim (reconfirmation) implementing

procedures which are in accordance with the current fuel licensing

criteria. The SLMCPRs ensure that greater than 99.9% of all fuel

rods in the core will avoid transition boiling if the limit is not

violated, thereby preserving the fuel cladding integrity. Therefore,

the proposed TS changes do not involve a reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

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

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

PA 17105.

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

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

PA 19101

NRC Project Director: John F. Stolz

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: January 15, 1997

Description of amendment request: The amendment proposes to

relocate the snubber operability, surveillance, and record requirements

for components (snubbers) in the Technical Specifications (TS) to plant

controlled documents.

Basis for proposed no significant hazards 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, based on the following:

1. These changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated

because:

The changes relocate operability, surveillance, and record

requirements for components (snubbers) which do not meet the

criteria for inclusion in the Technical Specifications (TS). The

affected components are not assumed to be initiators of analyzed

events and are not assumed to mitigate accident or transient events.

The snubber requirements will be relocated from the TS to plant

controlled documents. These requirements will be maintained pursuant

to 10 CFR 50.59. Therefore, the changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

different type of accident previously evaluated because:

The changes do not necessitate a physical alteration of the

plant (no new or different type of equipment will be installed) or

affect parameters governing normal plant operation. Adequate control

of future changes to snubber requirements will be maintained. Thus,

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

kind of accident from any accident previously evaluated for the

plant.

3. The proposed changes do not involve a reduction in a margin

of safety because:

The changes do not involve a change to the operability,

surveillance, and record requirements for the snubber program as

they currently exist in the TS, nor do they impact on any safety

analysis assumptions. The proposed changes relocate snubber

requirements from the TS to plant controlled documents. Changes to

the requirements in these documents are subject to the requirements

of 10 CFR 50.59. In addition, exceptions to code requirements for

testing will require NRC approval. Regulations and

[[Page 4353]]

FitzPatrick commitments to the NRC contain the necessary

programmatic requirements for the plant controlled documents.

Operating limitations will continue to be imposed, and required

surveillances will continue to be performed in accordance with

regulations, FitzPatrick commitments to the NRC, and written

procedures and instructions that are auditable by the NRC. If

snubber inoperability causes a TS system or component to be

inoperable, then the affected system or component Limiting Condition

for Operation (LCO) will be entered. Based on the above, 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: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa, Acting Director

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: January 7, 1997

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.2.5 to incorporate an exception

to the provisions of TS 4.0.4 and to clarify the time at which the

surveillance can be performed by adding that the surveillance is to be

performed within 24 hours after attaining steady state conditions at or

above 90% rated thermal power. The revised surveillance would also

contain editorial enhancements that do not change the intent of the

current surveillance. TS Table 3.2-1 for Salem Unit 1 would be revised

to delete reference to three loop operation (which is not permitted at

Salem Unit 1) in order to eliminate potential confusion when applying

this table.

Basis for proposed no significant hazards 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 changes proposed on the RCS [Reactor Coolant System] flow

measurement and exemption to Specification 4.0.4 do not affect the

operation of the equipment during conditions when they are required

to perform their safety function. No physical changes to the plant

result from the proposed changes made to the surveillance

requirements. The measurement of RCS flow does not impact the

probability of an accident.

Testing is being performed with the plant in the condition in

which the automatic initiation signals for low RCS flow would result

in a time consistent with the TS requirements.

Protection System in providing a reactor trip upon a loss of RCS

flow. Degradations in flow will occur over a long duration; however,

testing will continue to be performed within twenty-four hours upon

achieving steady state greater than or equal to 90% RTP [Rated

Thermal Power] after refueling which is a sufficiently short

duration after startup to identify flow degradations.

Changes proposed to refer to Table 3.2-1 for the DNB [Departure

from Nucleate Boiling] parameters and to delete the Unit 1 three

loop operation parameters, and the inclusion of the type of test

performed are editorial in nature.

Therefore, the consequences of an accident previously evaluated

are not significantly increased by the proposed changes.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not involve any modifications to

existing plant equipment, do not alter the function of any plant

systems, do not introduce any new operating configurations or new

modes of plant operation, nor change the safety analyses. The point

at which RCS flow is measured using a heat balance will not impact

the ability to maintain or monitor Reactor Coolant flows. The

proposed changes will, therefore, not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

Changes proposed to refer to Table 3.2-1 for the DNB parameters

and to delete the Unit 1 three loop operation parameters, and the

inclusion of the type of test performed are editorial in nature.

[The proposed changes will, therefore, not create the

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

previously evaluated.]

3. The proposed change does not involve a significant reduction

in a margin of safety.

The changes to the RCS flow surveillance do not decrease the

scope of the existing testing, but will clarify the point at which

the testing is performed.

The time in which testing is performed, after achieving steady

state conditions after reaching greater than or equal to 90% RTP

ensures that testing is performed in a timely manner. Flow margins

established as a result of previous testing will not be

significantly reduced in light of recent outage activities. Future

changes that might impact margins established by the testing will be

reviewed in accordance with the requirements of 10 CFR 50.59.

Changes proposed to refer to Table 3.2-1 for the DNB parameters

and to delete the Unit 1 three loop operation parameters, and the

inclusion of the type of test performed are editorial in nature.

All changes are consistent with the intent of Salem's current TS

[Technical Specification] and with the 18 month surveillances

specified in NUREG-1431, Revision 1.

The proposed change, therefore, 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: Salem Free Public library, 112

West Broadway, Salem, NJ 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

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph

M. Farley Nuclear Plant, Unit 1, Houston County, Alabama

Date of amendments request: December 26, 1996

Description of amendments request: The proposed amendment would

revise Technical Specification 3/4.4.6 ``Steam Generators'' and its

associated Bases. Specifically, the steam generator repair limit would

be modified to clarify that the appropriate method for determining

serviceability for tubes with outside diameter stress corrosion

cracking at the tube support plate is by a methodology that more

reliably assesses structural integrity. This amendment request is in

accordance with NRC's Generic Letter 95-05, ``Voltage-Based Repair

Criteria for Westinghouse Steam Generator Tubes Affected by Outside

Diameter Stress Corrosion Cracking.''

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

1) Operation of Farley units in accordance with the proposed

license amendment does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Testing of model boiler specimens for free standing tubes at

room temperature conditions shows burst pressures as high as

approximately 5000 psi for indications of outer diameter stress

corrosion cracking with voltage measurements as high as 26.5 volts.

Burst testing performed on pulled tubes,

[[Page 4354]]

including tubes pulled from Farley Unit 1, with up to 7.5 volt

indications show burst pressures in excess of 5300 psi at room

temperature. ... [T]ube burst criteria are inherently satisfied

during normal operating conditions by the presence of the tube

support plate. Furthermore, correcting for the effects of

temperature on material properties and minimum strength levels (as

the burst testing was done at room temperature), tube burst

capability significantly exceeds the R.G. [Regulatory Guide] 1.121

criterion requiring the maintenance of a margin of 1.43 times the

steam line break pressure differential on tube burst if through-wall

cracks are present without regard to the presence of the tube

support plate. Considering the existing data base, this criterion is

satisfied with bobbin coil indications with signal amplitudes over

twice the 2.0 volt voltage-based repair criteria, regardless of the

indicated depth measurement. This structural limit is based on a

lower 95% confidence level limit of the data at operating

temperatures. The 2.0 volt criterion provides an extremely

conservative margin of safety to the structural limit considering

expected growth rates of outside diameter stress corrosion cracking

at Farley. Alternate crack morphologies can correspond to a voltage

so that a unique crack length is not defined by a burst pressure to

voltage correlation. However, relative to expected leakage during

normal operating conditions, no field leakage has been reported from

tubes with indications with a voltage level of under 7.7 volts for a

3/4 inch tube with a 10 volt correlation to 7/8 inch tubing (as

compared to the 2.0 volt proposed voltage-based tube repair limit).

Thus, the proposed amendment does not involve a significant increase

in the probability or consequences of an accident.

Relative to the expected leakage during accident condition

loadings, the accidents that are affected by primary-to-secondary

leakage and steam release to the environment are Loss of External

Electrical Load and/or Turbine Trip, Loss of All AC Power to Station

Auxiliaries, Major Secondary System Pipe Failure, Steam Generator

Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a

Control Rod Drive Mechanism Housing. Of these, the Major Secondary

System Pipe Failure is the most limiting for Farley in considering

the potential for off-site doses. The offsite dose analyses for the

other events which model primary-to-secondary leakage and steam

releases from the secondary side to the environment assume that the

secondary side remains intact. The steam generator tubes are not

subjected to a sustained increase in differential pressure, as is

the case following a steam line break event. This increase in

differential pressure is responsible for the postulated increase in

leakage and associated offsite doses following a steam line break

event. In addition, the steam line break event results in a bypass

of containment for steam generator leakage. Upon implementation of

the voltage-based repair criteria, it must be verified that the

expected distributions of cracking indications at the tube support

plate intersections are such that primary-to-secondary leakage would

result in site boundary dose within the current licensing basis.

Data indicate that a threshold voltage of 2.8 volts could result in

through-wall cracks long enough to leak at steam line break

conditions. Application of the proposed repair criteria requires

that the current distribution of a number of indications versus

voltage be obtained during the refueling outages. The current

voltage is then combined with the rate of change in voltage

measurement and a voltage measurement uncertainty to establish an

end of cycle voltage distribution and, thus, leak rate during steam

line break pressure differential. The leak rate during a steam line

break is further increased by a factor related to the probability of

detection of the flaws. If it is found that the potential steam line

break leakage for degraded intersections planned to be left in

service coupled with the reduced allowable specific activity levels

result in radiological consequences outside the current licensing

basis, then additional tubes will be plugged or repaired to reduce

steam line break leakage potential to within the acceptance limit.

Thus, the consequences of the most limiting design basis accident

are constrained to present licensing basis limits, and therefore

there is no change to the probability or consequences of an accident

previously evaluated.

2) The proposed license amendment does not create the

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

previously evaluated.

Implementation of the proposed voltage-based tube repair

criteria does not introduce any significant changes to the plant

design basis. Use of the criteria does not provide a mechanism that

could result in an accident outside of the region of the tube

support plate elevations. Neither a single or multiple tube rupture

event would be expected in a steam generator in which the repair

criteria have been applied during all plant conditions. The bobbin

probe signal amplitude repair criteria are established such that

operational leakage or excessive leakage during a postulated steam

line break condition is not anticipated. Southern Nuclear has

previously implemented a maximum leakage limit of 140 gpd per steam

generator. The R.G. 1.121 criterion for establishing operational

leakage limits that require plant shutdown are based upon leak-

before-break considerations to detect a free span crack before

potential tube rupture. The 140 gpd limit provides for leakage

detection and plant shutdown in the event of the occurrence of an

unexpected single crack resulting in leakage that is associated with

the longest permissible crack length. R.G. 1.121 acceptance criteria

for establishing operating leakage limits are based on leak-before-

break considerations such that plant shutdown is initiated if the

leakage associated with the longest permissible crack is exceeded.

The longest permissible crack is the length that provides a factor

of safety of 1.43 against bursting at steam line break pressure

differential. A voltage amplitude of approximately 9 volts for

typical outside diameter stress corrosion cracking corresponds to

meeting this tube burst requirement at the 95% prediction interval

on the burst correlation. Alternate crack morphologies can

correspond to a voltage so that a unique crack length is not defined

by the burst pressure versus voltage correlation. Consequently, a

typical burst pressure versus through-wall crack length correlation

is used below to define the ``longest permissible crack'' for

evaluating operating leakage limits.

The single through-wall crack lengths that result in tube burst

at 1.43 times steam line break pressure differential and steam line

break conditions are about 0.54 inch and 0.84 inch, respectively.

Normal leakage for these crack lengths would range from about 0.4

gallons per minute to 4.5 gallons per minute, respectively, while

lower 95% confidence level leak rates would range from about 0.06

gallons per minute to 0.6 gallons per minute, respectively.

An operating leak rate of 140 gpd per steam generator has been

implemented. This leakage limit provides for detection of 0.4 inch

long cracks at nominal leak rates and 0.6 inch long cracks at the

lower 95% confidence level leak rates. Thus, the 140 gpd limit

provides for plant shutdown prior to reaching critical crack lengths

for steam line break conditions at leak rates less than a lower 95%

confidence level and for three times normal operating pressure

differential at less than nominal leak rates.

Considering the above, the implementation of voltage-based

repair criteria will not create the possibility of a new or

different kind of accident from any previously evaluated.

3) The proposed license amendment does not involve a significant

reduction in margin of safety.

The use of the voltage-based repair criteria is demonstrated to

maintain steam generator tube integrity commensurate with the

requirements of Generic Letter 95-05 and R.G. 1.121. R.G. 1.121

describes a method acceptable to the NRC staff for meeting GDC

[General Design Criteria] 2, 14, 15, 31, and 32 by reducing the

probability of the consequences of steam generator tube rupture.

This is accomplished by determining the limiting conditions of

degradation of steam generator tubing, as established by inservice

inspection, for which tubes with unacceptable cracking should be

removed from service. Upon implementation of the criteria, even

under the worst case conditions, the occurrence of outside diameter

stress corrosion cracking at the tube support plate elevations is

not expected to lead to a steam generator tube rupture event during

normal or faulted plant conditions. The most limiting effect would

be a possible increase in leakage during a steam line break event.

Excessive leakage during a steam line break event, however, is

precluded by verifying that, once the criteria are applied, the

expected end of cycle distribution of crack indications at the tube

support plate elevations would result in minimal, and acceptable

primary to secondary leakage during the event and, hence, help to

demonstrate radiological conditions are less than an appropriate

fraction of the 10 CFR [Part] 100 guideline.

The margin to burst for the tubes using the voltage-based repair

criteria is comparable to that currently provided by existing

technical specifications.

In addressing the combined effects of LOCA [loss-of-coolant

accident] + SSE [safe shutdown earthquake] on the steam generator

[[Page 4355]]

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

collapse may occur in the steam generators at some plants. This is

the case as the tube support plates may become deformed as a result

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

due to either the LOCA rarefaction wave and/or SSE loadings. Then,

the resulting pressure differential on the deformed tubes may cause

some of the tubes to collapse.

There are two issues associated with steam generator tube

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

RCS [reactor coolant system] flow area through the tubes. The

reduction in flow area increases the resistance to flow of steam

from the core during a LOCA which, in turn, may potentially increase

Peak Clad Temperature (PCT). Second, there is a potential the

partial through-wall cracks in tubes could progress to through-wall

cracks during tube deformation or collapse or that short through-

wall indications would leak at significantly higher leak rates than

included in the leak rate assessments.

Consequently, a detailed leak-before-break analysis was

performed and it was concluded that the leak-before-break

methodology (as permitted by GDC 4) is applicable to the Farley

reactor coolant system primary loops and, thus, the probability of

breaks in the primary loop piping is sufficiently low that they need

not be considered in the structural design basis of the plant.

Excluding breaks in the RCS primary loops, the LOCA loads from the

large branch line breaks were analyzed at Farley and were found to

be of insufficient magnitude to result in steam generator tube

collapse or significant deformation.

Regardless of whether or not leak-before-break is applied to the

primary loop piping at Farley, any flow area reduction is expected

to be minimal (much less than 1%) and PCT margin is available to

account for this potential effect. Based on analyses' results, no

tubes near wedge locations are expected to collapse or deform to the

degree that secondary to primary in-leakage would be increased over

current expected levels. For all other steam generator tubes, the

possibility of secondary-to-primary leakage in the event of a LOCA +

SSE event is not significant. In actuality, the amount of secondary-

to-primary leakage in the event of a LOCA + SSE is expected to be

less than that originally allowed, i.e., 500 gpd per steam

generator. Furthermore, secondary-to-primary in-leakage would be

less than primary-to-secondary leakage for the same pressure

differential since the cracks would tend to tighten under a

secondary-to-primary pressure differential. Also, the presence of

the tube support plate is expected to reduce the amount of in-

leakage.

Addressing the R.G. 1.83 considerations, implementation of the

tube repair criteria is supplemented by 100% inspection requirements

at the tube support plate elevations having outside diameter stress

corrosion cracking indications, reduced operating leakage limits,

eddy current inspection guidelines to provide consistency in voltage

normalization, and rotating probe inspection requirements for the

larger indications left in service to characterize the principle

degradation mechanism as outside diameter stress corrosion cracking.

As noted previously, implementation of the voltage-based repair

criteria will decrease the number of tubes that must be taken out of

service with tube plugs or repaired. The installation of steam

generator tube plugs or tube sleeves would reduce the RCS flow

margin, thus implementation of the voltage-based repair criteria

will maintain the margin of flow that would otherwise be reduced

through increased tube plugging or sleeving.

Considering the above, it is concluded that the proposed change

does not result in a significant reduction in margin with respect to

plant safety as defined in the Final Safety Analysis Report or any

bases of the plant Technical Specifications.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

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

Alabama 35201

NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph

M. Farley Nuclear Plant, Unit 1, Houston County, Alabama

Date of amendments request: January 10, 1997

Description of amendments request: The proposed amendments would

implement repair of tubes using laser welded tube sleeves for the steam

generators at Farley Units 1 and 2 as described in WCAP-13088, Revision

4, and WCAP-14740. In addition, for Unit 2, references to a one-cycle

limited implementation of L* are being removed. The approval for the

limited implementation of L* expired at the last Unit 2 outage in the

fall of 1996.

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

1. Operation of Farley Units 1 and 2 in accordance with the

proposed license amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The laser welded sleeve configurations as described within WCAP-

13088, Revision 4 and WCAP-14740 have been designed and analyzed in

accordance with the requirements of the ASME Code [American Society

of Mechanical Engineers Boiler and Pressure Vessel Code]. Fatigue

and stress analyses of the sleeved tube assemblies produced

acceptable results. Mechanical testing has shown that the structural

strength of Alloy 690 sleeves under normal, faulted and upset

conditions is within acceptable limits. Leakage testing for 7/8 inch

tube sleeves has demonstrated that significant primary-to-secondary

leakage is notexpected during all plant conditions, including the

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

tubesheet sleeve.

Initial acceptance of welded joints uses ultrasonic inspection

to verify that all weld thicknesses meet the minimum specified

conditions over the entire circumference. A plugging limit of 24%

allowable depth of penetration of the sleeve tube wall thickness

applies for each type of laser welded sleeve that may be installed

in the Farley Nuclear Plant steam generators and is determined for

uprated conditions with a limiting steam pressure for reduced

Thot and 20% steam generator tube plugging conditions. These

conditions represent the limiting primary-to-secondary operating

pressure differential, which is bounding for the sleeve plugging

limit and structural analysis inputs. However, the state-of-the-art

in eddy current inspection capability is such that no probes are

qualified to size the depth of penetration of stress corrosion

cracking. It is generally believed that the detection threshold of

these probes is well below 40% throughwall. Southern Nuclear

Operating Company will plug on detection any crack-like indications

that may occur in the sleeve using the sleeve inspection probe of

record until an inspection process is qualified to size depth of

penetration of stress corrosion cracking into the tube wall.

The hypothetical consequences of failure of the sleeve would be

bounded by the current steam generator tube rupture analysis

included in the Farley Nuclear Plant FSAR [Final Safety Analysis

Report]. Due to the slight reduction in diameter caused by the

sleeve wall thickness, it is expected that 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. Combinations of tubesheet sleeves and tube

support plate sleeves would reduce the primary fluid flow through

the sleeved tube assembly due to the series of diameter reductions

the fluid would have to pass on its way to the break area. The

overall effect would be reduced steam generator tube rupture release

rates.

As addressed previously, the proposed Technical Specification

change to support the installation of full length tubesheet,

elevated tubesheet, or tube support plate elevation Alloy 690 laser

welded sleeves as described in WCAP-13088, Revision 4 and WCAP-14740

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 Technical Specification minimum reactor coolant

[[Page 4356]]

system flow rate. The results of the analyses and testing, as well

as plant operating experience, demonstrate that the sleeve assembly

is an acceptable means of restoring tube integrity to a condition

consistent with its original design basis. Also, per Regulatory

Guide 1.83, Revision 1 recommendations, the condition of the sleeved

tube can be monitored through periodic inspections with present eddy

current techniques.

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 the installation of laser welded tube

sleeves will not increase the probability or consequences of an

accident previously evaluated. Depending upon the break location for

a postulated steam generator tube rupture event, implementation of

tube sleeving could act to reduce the radiological consequences to

the public due to reduced primary to secondary flow rate through a

sleeved tube compared to a non-sleeved tube based on the restriction

afforded by the sleeve wall thickness.

Removal of the references to the interim use of an L* repair

criteria will not involve a significant increase in the probability

or consequences of an accident previously evaluated.

2. The proposed license amendment does not create the

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

previously evaluated.

Implementation of laser welded sleeving will not introduce

significant or adverse changes to the plant design basis. Sleeving

also does not represent a potential to affect any other plant

component. Stress and fatigue analysis of the repair has shown the

ASME Code minimum stress values are not exceeded. Implementation of

laser welded sleeving 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 hypothetical 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. Since the sleeve

design does not affect any other component or location of the tube

outside of the immediate area repaired, in addition to the fact that

the installation of sleeves and the impact on current plugging level

analyses is accounted for, the possibility that laser welded

sleeving creates a new or different type of accident is not

credible.

Removal of the references to the interim use of an L* repair

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

of accident from any accident previously evaluated.

3. The proposed license amendment does not involve a significant

reduction in margin of safety.

The laser welded sleeving repair of degraded steam generator

tubes as identified in WCAP-13088, Revision 4, has been shown by

analysis to restore the integrity of the tube bundle consistent with

its original design basis condition as the requirements of the ASME

Code are satisfied. The safety factors used in the design of sleeves

for the repair of degraded tubes are consistent with the safety

factors in the ASME Boiler and Pressure Vessel Code used in steam

generator design. The design of the tubesheet sleeve lower joints

for the 7/8 inch sleeves (for both the full length and elevated

tubesheet sleeve) have been verified by testing to preclude

realistic leakage during normal and postulated accident conditions.

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 recommendations of Regulatory Guide 1.83, Revision 1

and the surveillance requirements included in Specification 4.4.6.0.

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-13088, Revision 4.

The effect of sleeving on the design transients and accident

analyses have been reviewed based on the installation of sleeves up

to the level of steam generator tube plugging coincident with the

minimum reactor flow rate. The installation of sleeves is to be

evaluated as the equivalent of some level of steam generator tube

plugging. Evaluation of the installation of sleeves is based on the

determination that LOCA evaluations for the licensed minimum reactor

coolant flow bound the effect of a combination of tube plugging and

sleeving up to an equivalent of the actual steam generator tube

plugging limit. Information provided in WCAP-13088, Revision 4,

describes the method to determine the flow equivalency for all

combinations of tubesheet and tube support plate sleeves in order

that the minimum flow requirements are met.

Implementation of laser welded sleeving will reduce the

potential for primary-to-secondary leakage during a postulated steam

line break while maintaining available primary coolant flow area in

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

tube through repair, primary pressure boundary integrity is restored

and the potential for primary-to-secondary leakage during all plant

conditions is minimized. These degraded tubes are returned to a

condition consistent with the design basis. While the installation

of a sleeve causes a reduction in primary coolant flow, the

reduction is significantly below the reduction incurred by plugging.

Therefore, greater primary coolant flow area is maintained through

sleeving.

Removal of the references to the interim use of an L* repair

criteria will not involve a significant 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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

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

Alabama 35201

NRC Project Director: Herbert N. Berkow

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear

Plant, Unit 1, Rhea County, Tennessee

Date of amendment request: January 10, 1997

Description of amendment request: The proposed amendment would

modify the Watts Bar Nuclear Plant (WBN) Unit 1 Technical

Specifications (TS) in order to implement the 1995 rule change to 10

CFR Part 50, Appendix J. The revised Appendix J provided an Option B

which allows performance based testing for containment leakage rate

testing. The TS in Section 3.6 and associated Bases, TS Section 3.0.2

and TS Section 5.7 would be changed. Also, the schedular exemption for

containment airlock testing now specified in the facility license in

Section 2.D(1) would no longer be required and would be deleted.

Basis for proposed no significant hazards determination: As

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

the issue of no significant hazards consideration, which is presented

below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendment to WBN TSs is in accordance with Option B

to 10 CFR 50, Appendix J. The proposed amendment adds a voluntary

performance-based option for containment leak-rate testing. The

changes being proposed do not affect the precursor for an accident

or transient analyzed in Chapter 15 of WBN Final Safety Analysis

Report. The proposed change does not increase the total allowable

primary containment leakage rate. The proposed change does not

reflect a revision to the physical design and/or operation of the

plant. [T]herefore, operation of the facility, in accordance with

the proposed change, does not significantly affect the probability

or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment to WBN TSs is in accordance with the new

performance-based option (Option B) to 10 CFR 50,

[[Page 4357]]

Appendix J. The changes being proposed will not change the physical

plant or the modes of operation defined in the facility license. The

proposed changes do not increase the total allowable primary

containment leakage rate. The changes do not involve the addition or

modification of equipment, nor do they alter the design or operation

of plant systems. Therefore, operation of the facility in accordance

with the propsoed change does not create the possibility of a new or

diferent kind of accident from any previously evaluated.

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

amendment would not involve a significant reduction in margin of

safety.

The proposed change to WBN TSs is in accordance with the new

option to 10 CFR 50, Appendix J. The proposed option is formulated

to adopt performance-based approaches. This option removes the

current prescriptive details from the TS. The proposed changes do

not affect plant safety analyses or change the physical design or

operation of the plant. The proposed change does not increase the

total allowable primary containment leakage rate. Therefore,

operation of the facility, in accordance with the proposed change,

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

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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

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

Date of amendment request: November 26, 1996, as supplemented

December 17, 1996

Description of amendment request: The proposed amendments would

allow a one-time only change necessary to replace the existing 125-volt

dc battery cells with new cells. Date of publication of individual

notice in Federal Register: December 13, 1996 (61 FR 65605)

Expiration date of individual notice: January 13, 1997

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina 28223-0001

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.

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

Unit No. 2, Pope County, Arkansas

Date of application for amendment: August 23, 1996

Brief description of amendment: This amendment makes Technical

Specifications changes allowing fuel enrichment of up to 5.0 weight

percent Uranium-235. The previous limit was 4.1 weight percent. This

change allows Arkansas Nuclear One, Unit-2, to receive, store, and use

nuclear fuel of 5.0 weight percent Uraninum-235.

Date of issuance: January 14, 1997

Effective date: January 14, 1997

Amendment No.: 178

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

Technical Specifications.

Date of initial notice in Federal Register: October 9, 1996 (61 FR

52964) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 14, 1997.No significant

hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Unit No. 2, Pope County, Arkansas

Date of application for amendment: November 24, 1996, as

supplemented on December 2, 1996.

Brief description of amendment: This amendment adds small break-

loss-of coolant accident methodology CENPD-137, Supplement 1-P and its

approval letter by the NRC as a reference to Section 6.9.5.1. This code

previously approved by the NRC increases the steam generator tube

plugging limit to 30% with an associated reduction of 10% in RCS flow.

This amendment also corrects a typographical error in Specification

6.9.5.1.8, and Specifications 6.9.5.1.10 through 6.9.5.1.14 are

numbered to accommodate these changes.

Date of issuance: January 14, 1997

Effective date: January 14, 1997

Amendment No.: 179

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

Technical Specifications.

Date of initial notice in Federal Register: December 3, 1996 (61 FR

64173) However, on December 9, 1996,

[[Page 4358]]

the licensee verified that the number of plugged tubes would not exceed

their current 10% limit established by the old code. This determination

removed the basis for considering this request as exigent. Since the

potential does exist for the plugging to exceed the 10% in the future,

the technical specification amendment request is therefore, a valid

request on a normal schedule. This change did not alter the staff's

initial proposed no safety hazard condition determination, therefore

noticing was not warranted. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated January 14, 1997.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station,

Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: October 24, 1996

Brief description of amendment: The amendment revises the technical

specifications to delete the accelerated testing requirements for the

standby diesel generators. This action is consistent with the

provisions of Generic Letter 94-01, ``Removal of Accelerated Testing

and Special Reporting Requiremets for Emergency Diesel Generators,''

dated May 31, 1994.

Date of issuance: January 14, 1997

Effective date: January 14, 1997

Amendment No.: 90

Facility Operating License No. NPF-47. The amendment revised the

Technical Specifications/operating license.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64384) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 14, 1997.No significant

hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station,

Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: May 30, 1996

Brief description of amendment: The amendment revises the technical

specification surveillance requirement 3.8.3.4 to specify a 5-start

pressure for the air recievers associated with the Division III, High

Pressure Core Spray emergency diesel generator.

Date of issuance: January 16, 1997

Effective date: January 16, 1997

Amendment No.: 91

Facility Operating License No. NPF-47. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34892) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 16, 1997.No significant

hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: September 23, 1996

Brief description of amendment: Changes to Technical Specification

(TS) to delete a note for the Surveillance Requirement 3.3.7.1 for the

Engineered Safeguard Actuation System Logic.Date of issuance: January

6, 1997

Effective date: January 6, 1997

Amendment No.: 155

Facility Operating License No. DPR-72. Amendment revised the TS.

Date of initial notice in Federal Register: October 23, 1966 (61 FR

55034) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 6, 1997.No significant

hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

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

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

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

Matagorda County, Texas

Date of amendment request: October 23, 1996, as supplemented by

letter dated November 6, 1996.

Brief description of amendments: The amendments revised Technical

Specification 3.4.6.1, regarding reactor coolant system leakage

detection instrumentation, to adopt the requirements found in NUREG-

1431, ``Standard Technical Specifications Westinghouse Plants,'' for

the reactor coolant system leakage detection instrumentation.

Date of issuance: January 8, 1997

Effective date: January 8, 1997

Amendment Nos.: Unit 1 - Amendment No. 86; Unit 2 - Amendment No.

73

Facility Operating License Nos. NPF-76 and NPF-80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64387) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 8, 1997.No significant

hazards consideration comments received: No

Local Public Document Room location: Wharton County Junior College,

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

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

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

Illinois

Date of application for amendment: April 19, 1996, and as

supplemented on August 15, 1996

Brief description of amendment: The amendment introduces new

Technical Specification (TS) 3.10.10, ``Single Control Rod Withdrawal -

Refueling,'' under TS 3.10, ``SPECIAL OPERATIONS.'' The purpose of this

Special Operations LCO is to permit the withdrawal of a single control

rod for testing in MODE 5 without imposing the requirements for

establishing the secondary containment and main control room boundaries

as normally required during Core Alterations.

Date of issuance: January 13, 1997

Effective date: January 13, 1997

Amendment No.: 112

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

Technical Specifications.

Date of initial notice in Federal Register: May 22, 1996 (61 FR

25707) and September 25, 1996 (61 FR 50344). The August 15, 1996,

submittal changed the focus of the original amendment request,

therefore, it was re-noticed in the FEDERAL REGISTER. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated January 13, 1997.No significant hazards consideration comments

received: No

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

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

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

Connecticut

Date of application for amendment: September 12, 1995

[[Page 4359]]

Brief description of amendment: The amendment revises Technical

Specification 6.3.1 to add a requirement that the Assistant Operations

Manager hold a senior reactor operator (SRO) license if the Operations

Manager does not hold an SRO license for Millstone Unit 3.

Date of issuance: January 7, 1997

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

within 60 days.

Amendment No.: 132

Facility Operating License No. NPF-49. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 27, 1996 (61 FR

13530) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 7, 1997.No significant

hazards consideration comments received: No.

Local Public Document Room location: 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

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 application for amendment: April 26, 1996, as supplemented

August 23, 1996.

Brief description of amendment: The amendment changes requirements

regarding reactor coolant system leakage testing following refueling

outage and other sytem pressure testing of reactor coolant system

following repairs, replacements, or modifications.

Date of issuance: January 7, 1997

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

60 days.

Amendment No.: 171

Facility Operating License No. DPR-64: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 5, 1996 (61 FR

28602) The August 23, 1996, letter provided clarifying information that

did not change the initial no significant hazards consideration

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

contained in a Safety Evaluation dated January 7, 1997No significant

hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: October 29, 1996

Brief description of amendment: The amendment corrects an error

with respect to Table 3.3.2-1, Function 6c of the Technical

Specifications (TSs) which references the incorrect Required Action for

inoperable channels of the auxiliary feedwater pump actuation on Steam

Generator Level - Low Low logic. The TSs are revised to correct the

Required Action to place the inoperable channel in ``trip'' within 6

hours or initiate a plant shutdown to Mode 4.

Date of issuance: January 9, 1997

Effective date: January 9, 1997

Amendment No.: 66

Facility Operating License No. DPR-18: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64395) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 9, 1997No significant

hazards consideration comments received: No

Local Public Document Room location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: October 29, 1996

Brief description of amendment: This amendment revises the MODE of

applicability for the motor-driven auxiliary feedwater pump actuation

on opening of the main feedwater pump breakers to correct an error

introduced during Amendment No. 61.

Date of issuance: January 9, 1997

Effective date: January 9, 1997

Amendment No.: 67

Facility Operating License No. DPR-18: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64395) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 9, 1997No significant

hazards consideration comments received: No

Local Public Document Room location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,

Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of application for amendments: April 29, 1996, as supplemented

October 21, December 2, and December 16, 1996

Brief description of amendments: These amendments revise Technical

Specification (TS) Section 15.3.14, ``Fire Protection System,'' and

Section 15.4.15, ``Fire Protection System,'' and relocate the

requirements of the fire protection program from the TS and

incorporate, by reference, the NRC-approved fire protection program

into the Final Safety Analysis Report. In addition, the amendments

revise the operating licenses to include the NRC's standard fire

protection condition. The amendments also approve administrative

changes consistent with the relocation as well as corrections to

several typographical errors.

Date of issuance: January 8, 1997

Effective date: January 8, 1997, and implementation within 90 days

from the date of issuance. Implementation shall include the relocation

of Technical Specification requirements to the appropriate licensee-

controlled document as identified in the licensee's application dated

April 29, 1996, as supplemented October 21, December 2, and December

16, 1996, and reviewed in the staff's safety evaluation dated January

8, 1997.

Amendment Nos.: Unit 1 - 170, Unit 2 - 174

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revise the Technical Specifications and the operating licenses.

Date of initial notice in Federal Register: June 5, 1996 (61 FR

28621) The October 21, December 2, and December 16, 1996, supplements

provided corrected license and TS pages and a 90-day implementation

schedule. These supplements were within the scope of the original

application and did not change the staff's initial proposed no

significant hazards considerations determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated January 8, 1997.No significant hazards consideration

comments received: No

Local Public Document Room location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: October 31, 1996

[[Page 4360]]

Brief description of amendment: The amendment revises Kewaunee

Nuclear Power Plant Technical Specification 6.9, ``Reporting

Requirements,'' by deleting the annual requirement to submit a

description of changes made pursuant to 10 CFR 50.59. Administrative

changes are also made to correct inconsistencies in the TS Table of

Contents and in a footnote for Table TS 3.5-1.

Date of issuance: January 6, 1997

Effective date: January 6, 1997

Amendment No.: 131

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64397) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 6, 1997.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

Dated at Rockville, Maryland, this 22nd day of January 1997.

For The Nuclear Regulatory Commission

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects - III/IV, Office of

Nuclear Reactor Regulation

[Doc. 97-1994 Filed 1-28-97; 8:45 am]

BILLING CODE 7590-01-F