Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating

LicensesInvolving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from October 11, 1996, through October 25, 1996.

The last biweekly notice was published on October 23, 1996.

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 December 6, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

persons should consult a current copy of 10 CFR 2.714 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

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

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

and (3) the possible effect of any order which may be entered in the

proceeding on the petitioner's interest. The petition should also

identify the specific aspect(s) of the subject matter of the proceeding

as to which petitioner wishes to intervene. Any person who has filed a

petition for leave to intervene or who has been admitted as a party may

amend the petition without requesting leave of the Board up to 15 days

prior to the first prehearing conference scheduled in the proceeding,

but such an amended petition must satisfy the specificity requirements

described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

by the above date. Where petitions are filed during the last 10 days of

the notice period, it is requested that the petitioner promptly so

inform the Commission by a toll-free telephone call to Western Union at

1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to (Project Director): petitioner's name

and telephone number, date petition was mailed, plant name, and

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert

Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

Date of amendment request: October 3, 1996

Description of amendment request: The proposed amendment changes

the provision for receiving, possessing and using byproducts, source

and special nuclear material at Calvert Cliffs Unit 1.

Currently, Unit 1 is licensed under 10 CFR Part 30 to receive,

possess, and use 100 millicuries of byproduct material for sample

analysis or instrument calibration, 500 millicuries of byproduct

material in the form of equipment; and 500 millicuries of Sodium-24 for

steam turbine acceptance testing. In addition, Unit 1 is licensed to

receive, possess and use 100 milligrams each of source or special

nuclear material under 10 CFR Parts 40 and 70. Unit 2 is licensed under

10 CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as

required any byproduct, source, or special nuclear material for sample

analysis or instrument calibration or associated with radioactive

apparatus or components. This proposed amendment would change the Unit

1 license to be consistent with the Unit 2 license by replacing license

conditions 2.B.3 and 2.B.4 with the same wording as Unit 2's license

condition 2.B.4.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident.

Currently, Unit 1 is licensed under 10 CFR Part 30 to receive,

possess, and use 100 millicuries of byproduct material for sample

analysis or instrument calibration, 500 millicuries of byproduct

material in the form of equipment; and 500 millicuries of Sodium-24

for steam turbine acceptance testing. Unit 1 is also licensed under

10 CFR parts 40 and 70 to receive, possess, and use 100 milligrams

of source or special nuclear material. Unit 2 is licensed under 10

CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as

required any byproduct, source, or special nuclear material for

sample analysis or instrument calibration or associated with

radioactive apparatus or components. This proposed amendment would

change the Unit 1 license to be consistent with the Unit 2 license.

The reason for this proposed change is that it is sometimes

necessary to receive and use byproduct material, sources, or special

nuclear material with different activity levels, and in different

quantities than is specified by the Unit 1 license.

The current licenses for the two units allow radioactive

materials to be accepted and used at Unit 2, although these same

materials would not be acceptable for use at Unit 1. These

byproduct, source, and special nuclear materials are used by the

same people and for the same function in either

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unit. Training and procedures for handling radioactive material have

been developed and used at both Units over the last 20 years. These

procedures are adequate to control the acceptance and use of

radioactive material at Unit 2 and, therefore, adequate to control

radioactive material at Unit 1.

Receiving, possessing, and using byproduct, source, or special

nuclear material is not related to accident conditions. Therefore,

changing the Unit 1 license conditions to be the same as the Unit 2

license condition does not involve a significant increase in the

probability or consequences of an accident.

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

of accident from any accident previously evaluated.

Procedures and training governing the acceptance and use of

radioactive materials are the same for both Unit 1 and Unit 2. These

procedures will not be changed as a result of this license change.

In addition, receiving, possessing, and using radioactive material

is not related to accident conditions. Therefore, making the Unit 1

license the same as the Unit 2 license will not create the

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

previously evaluated.

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

safety.

The margin of safety in this case is exposure to contaminated

material or equipment. Exposure is controlled by adequate training

and procedures. Radioactive material is received by personnel

assigned to the Radiation Safety Section. These personnel are

trained in receiving and shipping contaminated material. Once the

material is onsite, it becomes the responsibility of the radiation

protection staff who are trained in the handling of all levels of

radioactive material. Training and procedures for handling

radioactive materials have been developed and used over the 20-year

life of the plant, and are currently deemed adequate for compliance

with the Unit 2 license. Therefore, making the Unit 1 license the

same as the Unit 2 license will not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: S. Singh Bajwa, Acting Director

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

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

Date of amendment request: September 20, 1996

Description of amendment request: The proposed amendments would add

a footnote to specification 4.3.1.B.4.A.10.a which refers to a letter

that describes enhancements made to the Combustion Engineering sleeve

installation process.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendment continues to allow the Combustion

Engineering sleeves to be used as an alternate tube repair method

for Zion steam generators, along with the process enhancements which

are described in the letter identified in the proposed Technical

Specification note. The sleeve configuration, which was designed and

analyzed in accordance with the criteria of Regulatory Guide (RG)

1.121 and Section III of the ASME Code, is unaffected by the

enhancements. Fatigue and stress analyses of the sleeved tube

assemblies as described in the currently approved Topical Report,

CEN-331-P, Revision 1-P, are unaffected by the enhancements.

Mechanical testing which has shown that the structural integrity

of the sleeves under normal, faulted, and upset conditions is within

the acceptable limits and is unaffected by the enhancements. Leakage

rate testing for the tube sleeves which has demonstrated that

primary to secondary leakage is not expected during any plant

condition is unaffected by the enhancements. The consequences of

leakage through the sleeved region of the tube, including the

enhancements, is bounded by the existing steam generator tube

rupture (SGTR) analysis included in the Zion Updated Final Safety

Analysis Report.

The proposed Technical Specification change reflects

enhancements to the installation and inspection process identified

in Topical Report CEN-331-P, Revision 1-P, which is currently

referenced in the Technical Specifications. These enhancements do

not increase the probability or consequences of an accident

previously evaluated. The enhancement which disallows the

installation of the tube plugs made from Inconel 600 material was

done so based upon industry information and is addressed by NRC

Bulletin 89-01. The use of the Plus Point Probe, its associated data

acquisition equipment, and improved visual inspection equipment, are

conservative actions and improve the quality of the sleeving

process. The use of the mechanical plug in lieu of the welded plug

meets the established design requirements and is advantageous in the

area of dose reduction, because of reduced time to install. Minor

changes to the sleeve installation equipment as described in the

Topical Report, represent equipment enhancements and do not alter

the sleeve design or qualification testing.

The proposed Technical Specification change does not adversely

impact any previously evaluated design basis accident. Installation

of the sleeves, with the described enhancements, can be used to

repair degraded tubes by returning the condition of the tubes to

their original design basis condition for tube integrity and leak

tightness during all plant conditions. Therefore, the currently

approved sleeving process with the described enhancements will not

increase the probability of occurrence of an accident previously

evaluated.

Therefore, these proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The implementation of the enhancements to the proposed sleeving

process will not affect the plant design basis. The current stress

and fatigue analyses of the repair identified in Topical Report CEN-

331-P, Revision 1-P, has shown the ASME Code and RG 1.121 allowable

values are met and are unaffected by the described enhancements. The

current sleeving design, with the described enhancements, will

continue to maintain overall tube bundle structural integrity and

leak tightness at a level consistent with that of the originally

supplied tubing. Leak and mechanical testing of the sleeves, are

unaffected by the proposed enhancements and continue to support the

conclusions that the sleeve retains both structural integrity and

leak tightness during all operating and accident conditions. Repair

of a tube with a sleeve, utilizing the described enhancements, does

not provide a mechanism that results in an accident outside of the

area affected by the sleeve.

The described change to implement the cited enhancements will

not create a new or different type of accident. The change only

reflects enhancements to the currently approved installation/

inspection process and, would not change or impact any hypothetical

accident previously discussed. Use of improved Non-Destructive

Examination, data acquisition and visual inspection equipment

improves the quality of the sleeving process and has no negative

effect on the margin of safety. The elimination of the use of the

Inconel 600 plug also improves the margin of safety.

Any hypothetical accident as a result of potential tube or

sleeve degradation in the repaired portion of the tube is bounded by

the existing SGTR analysis. The sleeve design, including described

enhancements, does not affect any other component, or affect any

location on the tube outside of the immediate area repaired.

Therefore, the proposed changes do not create the possibility of

a new or different type of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

[[Page 57484]]

The currently approved sleeving repair of degraded steam

generator tubes has been shown by analysis to restore the integrity

of the tube bundle to its original design basis condition. By

implementing the described enhancements, the consistent quality of

the upper sleeve weld has increased thereby reducing the potential

for rework and reducing the potential for leaving a weld indication

in service.

The proposed change does not involve a reduction in the margin

of safety. The change reflects enhancements to the installation/

inspection processes which are currently referenced in the Technical

Specifications. These enhancements would not have any adverse

effects on the previously evaluated design transient or accident

analyses. The enhancements represent acceptable industry standards.

Therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

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

this review, it appears that the three standards of 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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

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

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

Date of amendment request: August 21, 1996

Description of amendment request: The proposed amendment will

modify Containment Penetrations Nos. 53 and 65 design by modifying the

design of instrumentation lines for Containment Vacuum Relief (CVR)

system that pass through these containment penetrations. The proposed

change will correct the error in previously docketed information that

was used by NRC during licensing process.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change will not increase the probability of

previously analyzed accidents. The proposed change seeks to clearly

document the design and licensing bases for acceptance of the CVR

sensing instrument lines. The proposed change to the monitoring

lines will provide greater assurance that containment integrity will

be maintained following a LOCA concurrent with a single active

failure. The design change to the non-essential monitoring line will

reduce the potential bypass leakage from penetrations 53 and 65 by

adding a redundant automatic containment isolation valve on

penetration 53 and isolating the non-essential instrument line on

penetration 65. This design change can be performed at power without

violating any license/regulatory requirements that ensure

containment integrity is maintained.

There is no change in the function of the instrumentation. The

only difference is that CVR-IDPT-5017B and C non-safety differential

transmitters that monitor the CVR system will be sensing containment

pressure from penetration 53. If the non-essential line coming from

penetration 53 becomes inoperable, containment to annulus

differential pressure can be obtained from alternate

instrumentation. The essential sensing line that actuates the CVR

system to protect containment within design vacuum pressure is not

affected by the design change.

Adding a redundant automatic containment isolation valve in

penetration 53's non-essential instrument line instead of the excess

flow check valve and isolating the non-essential line in penetration

65's will significantly reduce the potential bypass leakage. The

proposed change will credit the essential instrument lines as a

closed system outside containment. The appropriate testing and

acceptance criteria will be applied to ensure that any leakage

associated with these potential bypass leakage paths, will not

exceed the limits used in the Waterford 3 safety analysis or result

in a significant increase in analyzed dose consequences. Therefore,

the proposed change will not involve significant increase in the

probability or consequences of any accident previously evaluated.

The proposed change will credit the essential sensing lines

outside containment as a closed system and will not affect the plant

or the manner in which the plant [is] operated.

The failure modes associated with containment isolation remain

unchanged as a result of the design change to the non-essential

monitoring lines. The function of the non-safety instrumentation is

not affected. The only difference is that all of the non-safety

instrumentation will be sensing containment pressure from

penetration 53. However, if the non-essential line coming from

penetration 53 becomes inoperable, containment pressure can be

obtained from alternate instrumentation. Adding a redundant

automatic containment isolation valve in series with CVR 401A in the

non-essential instrument line ensures containment isolation

following a LOCA with a concurrent a single active failure.

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

new or different kind of accident from any accident previously

evaluated.

The addition of a redundant automatic containment isolation

valve in series with CVR 401A in the non-essential instrument line

breaching penetration 53 ensures containment isolation postulating a

single active failure on a Containment Isolation Actuation Signal

(CIAS). While this modification is performed, administrative

controls will require containment integrity to be maintained by a

seismic Category 1, ASME Section III, Class 2, passive containment

isolation device.

The essential CVR instrument sensing lines form a seismically

qualified, closed system outside containment which is designed for

pressure equal to or greater than containment. The instrument

cabinets C-3A(B) are seismic Category I and safety related. The

instruments are Safety Class 1E and have a static pressure rating of

1000 psig. These lines meet the criteria of BTP CSB 6-3 for

crediting a closed system as a leakage boundary to preclude bypass

leakage by being designed, fabricated, erected, and tested to

standards commensurate with the safety function to be performed. The

proposed change will apply the appropriate testing and acceptance

criteria to ensure that any leakage associated with these potential

bypass leakage paths, will not exceed the limits used in the

Waterford 3 safety analysis or result in a significant increase in

analyzed dose consequences. Therefore, the proposed change will not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: October 4, 1996 (TSCR No. 250)

Description of amendment request: The proposed Technical

Specification (TS) change reflects a change in the Safety Limit Minimum

Critical Power Ratio (SLMCPR) and as a result, a change in the

operating Minimum Critical Power Ratio limit.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

in the probability or consequences of an accident previously

evaluated.

The derivation of the revised SLMCPR for Oyster Creek for

incorporation into the TS,

[[Page 57485]]

and its use to determine cycle-specific thermal limits, have been

performed using NRC-approved methods. Additionally, interim

implementing procedures, which incorporate cycle-specific

parameters, have been used. Based on the use of these calculations,

the revised SLMCPR will not increase the probability or consequences

of an accident.

The basis of the MCPR Safety Limit 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 SLMCPR preserves the

existing margin to transition boiling and fuel damage in the event

of a postulated accident. The probability of fuel damage is not

increased.

Revising the operating MCPR limit for stability will ensure that

adequate margin is retained to the SLMCPR.

Therefore, the proposed TS change does 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 MCPR Safety Limit is a Technical Specification numerical

value designed to ensure that fuel damage from transition boiling

does not occur as a result of the limiting postulated accident. The

stability MCPR limit ensures an adequate operating MCPR margin to

the SLMCPR. These revised limits cannot create the possibility of

any new type of accident. The new SLMCPR has been calculated using

NRC-approved methods. Additionally, interim procedures, which

incorporate cycle-specific parameters, have been used. Therefore,

the proposed TS change does 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 SLMCPR is 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, have been used. The MCPR

Safety Limit remains high enough to 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 fuel cladding integrity.

The revised stability MCPR limit retains the existing margin to the

SLMCPR. Therefore, the proposed TS change does not involve a

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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

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

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

NRC Project Director: John F. Stolz

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: October 10, 1996 (TSCR No. 203)

Description of amendment request: The proposed Technical

Specification revision will extend the instrumentation surveillances

for Condenser Low Vacuum, High Temperature Main Steamline Tunnel,

Recirculation Flow, and Reactor Coolant Leakage. Additionally, the

change will extend the equipment tests/operability checks for

Containment Vent and Purge Isolation, Electromagnetic Relief Valve

Operability, and Drywell to Torus Leakage Test. The above change

extensions conform with the 24 month refueling interval.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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 extends the period between successive

refueling interval surveillance(s) to once every 24 months for those

surveillance(s) evaluated herein. The proposed surveillance interval

changes do not involve any change to the actual surveillance

requirements, nor does it involve any change to the limits and

restrictions on plant operations. The reliability of systems and

components relied upon to prevent or mitigate the consequences of

accidents previously evaluated is not degraded by the proposed

change to the surveillance interval. Assurance of system and

equipment availability is maintained. This change does not involve

any change to system or equipment configuration. Therefore, this

change does not increase the probability of occurrence or the

consequences of an accident previously evaluated.

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 extends the period between successive

refueling interval surveillance(s) to once every 24 months for those

surveillance(s) evaluated herein. The proposed surveillance interval

changes do not involve any change to the actual surveillance

requirements, nor does it involve any change to the limits and

restrictions on plant operation. This change does not involve any

change to system or equipment configuration. Therefore, this change

is unrelated to the possibility of creating a new or different kind

of accident from any previously evaluated.

Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendment extends the period between successive

refueling interval surveillance(s) to once every 24 months (+/-25%

or 30 months) for the surveillances evaluated herein. The proposed

surveillance interval changes do not involve any change to the

actual surveillance requirements, nor does it involve any change to

the limits and restrictions on plant operation. The reliability of

systems and components is not degraded by the proposed change to the

surveillance interval. Assurance of system and equipment

availability is maintained. Therefore, it is concluded that

operation of the facility in accordance with the proposed amendment

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

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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

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

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

NRC Project Director: John F. Stolz

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

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: October 10, 1996 (TSCR No. 243)

Description of amendment: The proposed Technical Specification (TS)

will change the trip setting for TS Table 3.1.1 Item G.3, Automatic

Depressurization System (ADS) by clarification of the functional

requirement to provide an interlock permissive which ensures that a

source of cooling water is available via the Core Spray System prior to

depressurization. This will be accomplished by replacing the present

interlock description ``AC Voltage'' with core spray booster pump

differential pressure, as the permissive required for initiation of

ADS. A corresponding surveillance requirement is being added to TS

Table 4.1.1 which reflects the need to test and calibrate the core

spray booster pump differential pressure switches pursuant to existing

[[Page 57486]]

plant procedures. Additionally, allowed outage time (AOT) is addressed

in the footnote ``i'' for the differential pressure switches based upon

the currently designed ADS logic trains and footnote ``h'' to parallel

the ``Low-Low Reactor Water Level'' and ``High Drywell Pressure'' AOTs

associated with Standard Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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 implementation of this TSCR does not involve an increase in

the probability of occurrence or the consequences of an accident

previously evaluated, as no plant modifications are proposed by the

change request, and no changes in instrument set or reset setpoints

are required in order to implement the change. This change serves to

clarify and to incorporate the ``as-built'' ADS system logic

parameter (core spray booster pump differential pressure) as the

functional permissive required for initiation of ADS. This

``interlock'' permissive compares closely with that of the BWR

[boiling-water reactor] STS [Standard Technical Specifications]

requirement to monitor core spray discharge pressure for initiation

of ADS. In addition, the AOTs for the ADS initiation signals are

being revised to align with the AOTs provided for such signals in

the STS. The performance and function of the Automatic

Depressurization System is unchanged by this request. However, by

implementation of the change the specific functions of the ADS as-

built d/p permissives would then be clearly identified in and

controlled by T.S. Table 3.1.1, ``Protective Instrumentation

Requirements,'' including the associated surveillance requirements

as shown on the revised T.S. Table 4.1.1.

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 implementation of this TSCR does not impact upon the

operation of the facility, and would not create the possibility of a

new or different kind of accident from any previously evaluated

because no plant modifications are proposed by this change request,

and no changes in instrument set or reset setpoints are required in

order to implement the change. This change clarifies the technical

specifications by incorporating the ``as-built'' ADS system logic

parameter (core spray booster pump differential pressure) as the

functional permissive required for initiation of ADS. This

``interlock'' permissive compares closely with that of the BWR STS

requirement to monitor core spay discharge pressure. The revised

AOTs for ADS initiation signals are also being changed to conform

with those allowed by and provided in the STS. The performance and

function of the Automatic Depressurization System (ADS) is unchanged

by this request.

OC plant surveillance procedures for both ADS and the Core Spray

system presently incorporate the calibration requirements and both

the set and reset setpoints calculated for the core spray booster

pump d/p switch permissive to the ADS initiation logic. Hence, a new

or different kind of accident from any previously evaluated is not

created.

Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The implementation of this TSCR does not involve a reduction in

the margin of safety for operation of the ADS or the Core Spray

system. The Technical Specification Bases which presently define the

margin of safety are not impacted as the core spray booster pump d/p

``interlock'' permissives are not described in the specifications

for ``Protective Instrumentation Requirements'' or its surveillance

requirements. In addition, the margin of safety for ADS initiation

is not reduced by this TSCR because the required system response is

not affected by the proposed changes as no plant modifications are

required which could create a potential impact upon the margins of

safety previously established.

The revision of AOTs associated with ADS actuation signals by

extension form 72 hours to 4 days is consistent with that presently

provided in the STS. This does not decrease the margin of safety

associated with availability of ADS as placement of the initiation

signals into the ``tripped condition'' maintains the operability of

the ADS trip systems while in the automatic mode. Additionally, the

Bases for STS Specification 3.1 provides justifications for AOTs

using the GE [General Electric] reliability analyses referenced

therein and therefore 4 days is both justified and conservative. The

margin of safety with respect to the instrument channels ability to

perform its intended actuation function is not impacted; therefore,

there is no reduction in the margin of safety.

Lastly, the surveillance frequency for the new surveillance

interval created on Table 4.1.1 for the d/p [s]witches is consistent

with that established in Reference 2 of the Bases for Technical

Specification 4.1. Therefore, there is no reduction in the margin of

safety as a result of this change request.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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

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

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

NRC Project Director: John F. Stolz

Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.

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

Date of amendment request: October 17, 1996

Description of amendment request: The proposed amendment would

revise Facility Operating License NPF-62 to acknowledge the transfer of

Soyland Power Cooperative's 13.21% minority ownership interest in the

Clinton Power Station to Illinova Power Marketing, Inc., the

unregulated power marketing affiliate of Illinois Power, and a wholly

owned subsidiary of Illinova Corporation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated because it merely revises the Operating License

to indicate the transfer of a minority ownership interest to the

corporate parent of the majority owner and licensee. This proposed

amendment represents an administrative rather than operational

change and, therefore, has no impact on accidents previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated because Illinois Power will continue to be the operator of

Clinton

Power Station, and further, there will be no change to the

plant's physical configuration or operating philosophy as a result

of this proposed amendment.

3. The proposed amendment does not involve a significant

reduction in the margin of safety because it is only an

administrative change and will have no impact on any margin of

safety related to the design or operation of the facility.

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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th Street,

Decatur, Illinois 62525

NRC Project Director: Gail H. Marcus

[[Page 57487]]

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

Atomic Power Station, Lincoln County, Maine

Date of amendment request: September 13, 1996

Description of amendment request: The proposed amendment would

revise the Maine Yankee containment testing technical specification (TS

4.4) to implement 10 CFR Part 50, Appendix J, Option B, by referring to

Regulatory Guide 1.163, ``Performance-Based Containment Leakage-Test

Program'' dated September 1995.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. This amendment request does not involve a significant

increase in the probability or consequences of an accident

previously evaluated, because the proposed changes to the Technical

Specification do not affect the assumption, parameters or results of

any FSAR accident analysis.[...] These changes potentially result in

a minor increase in the consequences of an accident previously

evaluated due to the increased testing intervals. However, the

proposed changes do not result in an increase in the probability of

an accident previously identified since the containment system is

used for mitigation purposes only. The changes are also expected to

result in increased attention to components with poor leakage test

history as part of the performance-based nature of Option B such

that the marginally increased consequences from the expanded testing

intervals may be further reduced or negated. The addition of the

''...[as modified by approved] exemptions'' phrase is an

administrative change. Any specific exemptions from the requirements

of Appendix J will continue to require a submittal under 10 CFR

50.12 and subsequent review and approval by the NRC prior to

implementation. Therefore, these changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. Operation of Maine Yankee in accordance with the proposed

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

of accident from any accident previously evaluated. The proposed

changes do not involve a physical alteration of the plant (i.e., no

new or different type of equipment will be installed) nor alter the

function of the containment system. The changes only provide for

additional time between leakage tests and an increase in the test

pressure value equal to the containment design pressure which bounds

the containment peak accident pressure. Thus, these changes do not

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

any previously evaluated.

3. Operation of Maine Yankee in accordance with the proposed

changes does not involve a significant reduction in a margin of

safety. The proposed changes do not alter the manner in which safety

limits, limiting safety system setpoints, or limiting conditions for

operation are determined. The changes are expected to result in an

increased focus on components demonstrating poor leakage test

history without excessive testing of components which continue to

demonstrate good test history. Therefore, these changes do not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration. Local

Public Document Room location: Wiscasset Public Library, High Street,

P.O. Box 367, Wiscasset, ME 04578

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

Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director:

John A. Zwolinski

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

YankeeAtomic Power Station, Lincoln County, Maine

Date of amendment request: September 13, 1996, as supplemented

September 25, 1996

Description of amendment request: The proposed amendment would

revise TS 5.5.B to eliminate references to the Vice President (YNSD)

and designate the President, Maine Yankee, as the responsible official

for matters related to the composition, review and audit

responsibilities, authority and recordkeeping responsibilities of the

Nuclear Safety Audit and Review (NSAR) Committee. Minor editorial

changes are also proposed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below.

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change is administrative in nature and will not

have a direct effect on the physical plant or the maintenance of the

physical plant. The audit and review functions of the NSAR Committee

will continue to be required. The proposed changes will not, of

themselves, decrease the effectiveness of these functions. This

authority and responsibility realignment will continue to assure

that NSAR Committee has direct access to a level of management

necessary to perform their audit and review functions.

Since, the proposed change will not adversely effect the audit

and review functions of the NSARC and since the proposed change will

not have a direct effect on the physical plant or maintenance of the

physical plant, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change is administrative in nature and does not

introduce any new structures, systems, or components into the plant

design. This change continues to ensure that the NSAR Committee

reports to a management level such that there is sufficient

authority and organizational freedom to execute their audit and

review functions. Consequently, an unbiased oversight of the

programs and procedures is not compromised by this proposed change.

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 change does not involve a significant reduction

in the margin of safety.

The proposed change realigns the authority and responsibility

relationship of the NSAR Committee. The NSAR Committee will continue

to maintain effective oversight of programs and procedures. The

proposed change will continue to ensure that the NSAR Committee is

sufficiently independent from cost and schedule when opposed to

safety considerations. Therefore, the proposed change does not

involve a significant reduction in the margin of safety.

Based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Wiscasset Public Library, High

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

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

Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director:

John A. Zwolinski

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,

Millstone Nuclear Power Station, Unit 1, New London County,

Connecticut

Date of amendment request: July 2, 1996

Description of amendment request: The proposed amendment

incorporates limiting conditions for operation and surveillance

requirements for the safety/relief valve (SRV) electrical lift design

modification. The proposed amendment also makes clarification and

editorial changes, as well as revising the associated Bases section.

[[Page 57488]]

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Pursuant to 10CFR50.92, NNECO has reviewed the proposed change

and concludes that the change does not involve a significant hazards

consideration (SHC) since the proposed change satisfies the criteria

in 10 CFR 50.92(c). That is, the proposed change does not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The safety relief valves are considered for two analyzed

accidents, an overpressure transient (such as MSIV [main steam

isolation valve] closure with flux scram) and an inadvertent SRV

opening.

The new technical specifications do not affect normal operation,

therefore, they cannot increase the probability of an overpressure

event. Since the mechanical function will not be affected by the new

equipment, the new LCOs [limiting conditions for operation], or the

new surveillance requirements, there is no adverse affect on the

consequences of an overpressure event. The SRVs will be expected to

lift mechanically. If they do not open at the design setpoints, the

electrical actuation, which has the same setpoints, will cause the

valves to open less than 400 milliseconds later.

Sufficient redundancy and diversity is established for the

electrical lift by the use of two sensors in a two-out-of-two-taken-

once configuration. Therefore, the failure of any single component

cannot result in an inadvertent opening of an SRV. The only proposed

surveillance performed while at power is the daily instrument check.

This surveillance does not require the manipulation of any controls

and, as such, cannot affect the probability of an accident.

Therefore, based on the above, the proposed change to the

Technical Specifications does not involve a significant increase in

the probability or consequences of any previously evaluated

accident.

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

from any accident previously evaluated.

None of the proposed new LCOs or surveillance requirements has a

potential for creating a new or different kind of accident.

Expanding the LCO and surveillance requirements to address both the

mechanical actuation and the pressure sensor lift does not change

the type of action that these valves are expected to perform, nor

does it change the initial ``as-left'' requirements for the valves.

Plant operating parameters have also not changed.

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

or different kind of accident from any accident previously

evaluated.

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

The margin of safety previously analyzed for the SRVs was based

on the current nominal setpoints and allowable percent drift. The

electrical lift system improves the confidence that the SRVs will

lift within the specified range. The setpoint uncertainty of the

electrical lift system is similar to the drift allowed for the

mechanical lift in the Technical Specifications. All existing

functions that may actuate the SRVs (safety, manual, or automatic

lift) remain unaffected. The design of the pressure transmitters,

combined with the logic configuration, minimizes the possibility of

inadvertently opening the SRVs.

Therefore, this change has no impact on the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385

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

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

CT 06141-0270

NRC Project Director: Phillip F. McKee

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

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

Pennsylvania

Date of amendment request: March 29, 1996

Description of amendment request: The proposed Technical

Specifications (TS) changes would revise TS Surveillance Requirement

(SR) 4.5.1.d.2.b to delete the requirement to perform in-situ

functional testing of the Automatic Depressurization System (ADS)

valves once every 24-months as part of start-up testing activities.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS change does not involve any physical changes to

plant structures, systems, or components (SSC). The ADS will

continue to function as designed. The ADS is an Emergency Core

Cooling System (ECCS) designed to mitigate the consequences of an

accident, and therefore, can not contribute to the initiation of any

accident. The ADS utilizes five (5) of the 14 main steam line SRVs

as the primary method for depressurizing the reactor pressure vessel

to permit low pressure core cooling capability in the event of a

small break Loss-of-Coolant-Accident (LOCA) if the high pressure

cooling systems (i.e., High Pressure Coolant Injection (HPCI) and

Reactor Core Isolation Cooling (RCIC) systems) fail to maintain

adequate reactor vessel water level.

Deleting the TS SR to perform the in-situ testing of the ADS/

SRVs during start-up, as proposed, should reduce the probability of

an inadvertent opening of an SRV as discussed in Section 15.1.4 of

the LGS Updated Final Safety Analysis Report (UFSAR) since deleting

this testing requirement will eliminate a known initiator of SRV

pilot leakage and subsequent erosion. This proposed TS change will

have a tendency to increase, rather than decrease, the reliability

of the ADS/SRVs by eliminating the in-situ ADS functional start-up

testing. The probability of the ADS/SRVs to open on demand has been

demonstrated to be extremely high and is not measurably improved

through the in-situ ADS functional start-up testing.

This proposed TS change will not increase the probability of

occurrence of a malfunction of any plant equipment important to

safety. Alternate testing methods at LGS, Units 1 and 2, and at the

off-site test facility, adequately demonstrate proper ADS valve

operation and assure that the valves will continue to function as

designed. Existing surveillance testing and inspections of the ADS/

SRVs at LGS verify that the ADS initiation logic, solenoid valve

operation, pneumatic gas supply integrity and air operator assembly

(including pilot rod) will operate as designed. Offsite testing

verifies pilot disc operation, setpoint calibration and main valve

disc operation.

Deleting the in-situ testing requirement, as proposed, will

reduce the probability of inflating SRV leakage which should reduce

the probability of an inadvertent SRV opening. It has been

documented throughout the BWR industry that pilot disc leakage leads

to pilot disc and rod erosion, which can ultimately result in an

inadvertent opening of an SRV. Therefore, any SRV pilot leakage that

can be eliminated would reduce the probability of occurrence of a

malfunction of that SRV.

Deleting the ADS/SRV in-situ functional test will in no way

increase any consequences of a malfunction of plant equipment

important to safety. The consequences of a malfunction of an ADS/SRV

as discussed in the LGS UFSAR remain unchanged.

In addition, eliminating a known initiator of SRV leakage, as

proposed in this TS change, would help to reduce operator

workarounds in the form of suppression pool cooling and letdown

operation activities. As a result, this will reduce the unnecessary

operation of the Residual Heat Removal (RHR) and Residual Heat

Removal Service Water (RHRSW) systems.

Therefore, the proposed TS change does not involve an increase

in the probability or consequences of an accident previously

evaluated.

[[Page 57489]]

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

new or different kind of accident from any accident previously

evaluated.

This proposed TS change does not involve any physical changes to

plant SSC. The design and operation of the ADS/SRVs is not changed

from that currently described in the Safety Analysis Report (SAR).

The ADS will continue to function as designed to mitigate the

consequences of an accident. No changes of any kind are being made

to the valves, auxiliary components, or ADS logic. Deleting the

requirement to perform the ADS in-situ functional test during plant

start-up as proposed in this TS Change Request reduces the

likelihood of a SRV developing a leak and degrading throughout the

subsequent operating cycle. There is no possibility that

implementing this proposed TS change would create a different type

of malfunction to the ADS/SRVs than any previously evaluated.

Eliminating the requirement to perform the in-situ testing of

the ADS/SRVs during start-up activities, does not create a new or

different type of accident than any previously evaluated. There is

no accident scenario associated with testing the ADS/SRVs other than

the inadvertent opening of a relief valve which is currently

discussed in Section 15.1.4 of the LGS UFSAR. This proposed TS

change does not alter the conclusions described in the UFSAR

regarding an inadvertent opening of an SRV. No new or different type

of accident will be created as a result of this proposed TS change.

Therefore, the proposed TS change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

The proposed TS change does not involve any physical changes to

plant SSC. The design and functional requirements of the ADS will

not change. The ADS will still function as designed to mitigate the

consequences of an accident.

This proposed TS change involves deleting the requirement to

perform in-situ functional testing of the ADS/SRVs during start-up

activities. This testing imposes an unnecessary challenge on the

ADS/SRVs and has been linked to SRV degradation (e.g., pilot valve

and/or main valve leakage). This proposed TS change should reduce

SRV leakage and improve ADS/SRV reliability by reducing the

potential for spurious SRV actuation. The LGS TS Bases do not

identify specific testing requirements for ADS. ADS operability can

be readily demonstrated with extremely high confidence by the

existing additional surveillance tests and inspections performed for

the ADS. There will be no reduction in any margin of safety

resulting from this proposed TS change.

Therefore, the proposed TS change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

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

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

Pennsylvania

Date of amendment request: August 8, 1996

Description of amendment request: The proposed Technical

Specifications (TS) changes would revise TS Sections 3/4.3.1, ``Reactor

Protection System Instrumentation,'' 3/4.3.2, ``Isolation Actuation

Instrumentation,'' 3/4.3.3, ``Emergency Core Cooling System Actuation

Instrumentation,'' and the associated TS Bases Sections 3/4.3.1 and 3/

4.3.2 to eliminate selected response time testing requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS changes do not make any physical alterations or

modifications to the plant systems or equipment. The proposed

changes do not affect the capability of the associated systems to

perform their intended functions within their required response

times, nor do the proposed changes adversely impact the operation of

any plant equipment. The affected plant systems will continue to

function as designed. Elimination of the response time testing

requirements as proposed by this TS change for selected components

in RPS Instrumentation, Isolation Actuation System Instrumentation,

and ECCS Actuation Instrumentation will not adversely affect the

operation of these components.

The supporting analysis provided in NEDO-32291, demonstrates

that response time testing is redundant to other TS required

testing. NEDO-32291 demonstrated that these other required tests

(i.e., channel checks, channel calibrations, channel functional

tests, and logic system functional tests), in conjunction with

actions taken in response to NRC Bulletin 90-01 and NRCB 90-01,

Supplement 1, are sufficient to identify failure modes or

degradation in instrument response times, and ensure operation of

the associated systems within acceptable limits. There are no known

failure modes that can be detected by response time testing that

cannot also be detected by other TS required testing. The continued

application of other existing TS required testing such as channel

checks, channel calibrations, channel functional tests, and logic

system functional tests, ensures that the response times for these

systems will be maintained within the acceptance limits. The

capability of these systems to perform their intended functions

within their required response times is not adversely impacted by

this proposed TS change. NEDO-32291 evaluated the potential failure

modes of the affected instrumentation loops which could impact the

instrument loop response times. Industry operating experience was

also reviewed to identify failures that affect response times and

how they are detected. The failure modes identified were evaluated

to determine if other TS required surveillances and actions taken in

response to NRC Bulletin 90-01, and NRCB 90-01, Supplement 1, would

detect any effects on response time. There are no failures [sic]

[failure] modes identified that can be detected by response time

testing that cannot also be detected by other TS required testing.

PECO Energy has confirmed the applicability of the generic

evaluation provided in NEDO-32291 to LGS, Units 1 and 2. By letter

dated December 28, 1994, the NRC concluded that response time

testing can be eliminated from the TS for the selected

instrumentation identified in NEDO-32291, with certain provisions,

and that NEDO-32291 can be referenced in license amendment requests.

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 proposed TS changes do not involve any physical changes to

plant systems or equipment. The proposed changes apply only to the

testing requirements for the selected components involved and do not

result in any physical modifications to these components, or to

other plant system components. Elimination of the response time

testing requirements as proposed by this TS change for selected

components in RPS Instrumentation, Isolation Actuation System

Instrumentation, and ECCS Actuation Instrumentation will not

adversely affect the operation of these components. These components

will continue to function as designed. Consequently, no new failure

modes are introduced as a result of the proposed TS changes.

Eliminating the response time testing requirements as proposed,

does not create a new or different type of accident than any

previously evaluated. No new or different type of accident will be

created as a result of this proposed TS change.

NEDO-32291 demonstrates that other required tests (i.e., channel

checks, channel calibrations, channel functional tests, and logic

system functional tests), in conjunction with actions taken in

response to NRC Bulletin 90-01 and NRCB 90-01, Supplement

[[Page 57490]]

1, are sufficient to identify failure modes or degradation in

instrument response times, and ensure operation of the associated

systems within acceptable limits. There are no known failure modes

that can be detected by response time testing that cannot also be

detected by other TS required testing, and therefore, response time

testing for the selected components is redundant to the other TS

required testing.

Therefore, the proposed TS changes do not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed TS changes do not involve any physical changes to

plant systems or equipment. The proposed TS changes do not affect

the capability of the associated systems or equipment from

performing their intended functions. The systems involved will

continue to respond within their allowed response times. Elimination

of the response time testing requirements are based on the

evaluation provided in NEDO-32291 which demonstrates that response

time degradation can be detected by other TS required testing. The

evaluation concluded that other TS required tests (i.e., channel

checks, channel calibrations, channel functional tests, and logic

system functional tests), in conjunction with actions taken in

response to NRC Bulletin 90-01 and NRCB 90-01, Supplement 1, are

sufficient to identify failure modes or degradation in instrument

response times, and ensure operation of the associated systems

within acceptable limits.

In addition, although not specifically evaluated, the proposed

TS changes will provide an improvement to plant safety and operation

by reducing the time safety systems are unavailable, reducing the

potential for safety system actuations, reducing plant operating and

shutdown risk, limiting radiation exposure to plant personnel, and

eliminating the diversion of key personnel to conduct unnecessary

testing. Therefore, PECO Energy considers that the proposed TS

changes will result in an overall increase in the margin of safety

and that the changes do not constitute an unreviewed safety

question.

Therefore, the proposed TS changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket No. 50-353, Limerick

Generating Station, Unit 2, Montgomery County, Pennsylvania

Date of amendment request: August 1, 1996

Description of amendment request: The proposed Technical

Specifications (TS) changes would revise TS Section 3/4.4.6 (i.e.,

Figure 3.4.6.1-1) to reflect the addition of two hydrotest curves,

effective for 6.5 and 8.5 Effective Full Power Years (EFPY), to the

existing Pressure-Temperature Operating Limit (PTOL) curves for LGS

Unit 2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed Technical Specification (TS) change includes

Pressure-Temperature Operating Limit (PTOL) curves which were

conservatively generated in accordance with the fracture toughness

requirements of 10CFR50, Appendix G. The Adjusted Reference

Temperatures to the initial nil ductility reference temperatures

(RTNDT) used to evaluate the pressure/temperature limits for the

beltline materials were based on Regulatory Guide 1.99, Revision 2.

Future analyses of the Reactor Pressure Vessel (RPV) surveillance

capsule contents and future revisions to the PTOL curve as required,

ensure that the reactor pressure boundary will behave in a non-

brittle manner during plant testing, startup, and operation

throughout the life of the plant. The current schedule for removal

of the surveillance specimens from Limerick Generating Station (LGS)

Unit 2 RPV is during 2R05. The proposed change does not impact the

existing PTOL curves for 10 Effective Full Power Years (EFPY),

currently shown in the LGS Unit 2 TS. The proposed change only

provides additional information (i.e., two new curves) related to

the RPV condition following 6.5 and 8.5 EFPY, in order to facilitate

hydrostatic testing performed after 2R04 and 2R05, respectively. The

added PTOL curves are established in compliance with the methodology

used to calculate the predicted irradiation effects on vessel

beltline materials as documented in the LGS Updated Final Safety

Analysis Report (UFSAR). There are no physical changes to the plant

being introduced by the added PTOL curves.

Therefore, the proposed (TS) change does not involve an increase

in the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed Technical Specification (TS) change includes

Pressure-Temperature Operating Limit (PTOL) curves which were

conservatively generated in accordance with the fracture toughness

requirements of 10CFR50, Appendix G. The Adjusted Reference

Temperatures to the initial nil ductility reference temperatures

(RTNDT) used to evaluate the pressure/temperature limits for the

beltline materials were based on Regulatory Guide 1.99, Revision 2.

The proposed changes do not impact the existing PTOL curves for 10

Effective Full Power Years (EFPY), currently shown in the TS. They

only provide additional information (i.e., two new curves) related

to the reactor pressure vessel condition for 6.5 and 8.5 EFPY, in

order to facilitate hydrostatic testing performed after 2R04 and

2R05, respectively. The added PTOL curves are established in

compliance with the previous methodology used to calculate the

predicted irradiation effects on vessel beltline materials as

documented in the LGS [Updated Final Safety Analysis Report] UFSAR.

The proposed TS change does not involve any physical changes to

safety-related equipment.

Therefore, the proposed TS change does not create the

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

accident previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

The proposed change to Technical Specifications (TS) does not

reduce the margin of safety as defined in the Bases for any TS. The

added Pressure-Temperature Operating Limit (PTOL) curves for 6.5 and

8.5 Effective Full Power Years (EFPY) corresponding to 2R04 and

2R05, respectively, have been calculated in accordance with the

existing methodology used to calculate the PTOL curves currently

existing in the LGS Unit 2 TS (i.e., complying with the requirements

of 10CFR50 Appendix G, and Regulatory Guide 1.99, Revision 2) and

will more closely reflect the actual required reactor pressure

vessel condition at the time in which the hydrotest is performed.

Therefore, the margin of safety is not affected.

Therefore, the proposed TS change does not involve a reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

[[Page 57491]]

Philadelphia Electric Company, Docket No. 50-353, Limerick

Generating Station, Unit 2, Montgomery County, Pennsylvania

Date of amendment request: August 5, 1996

Description of amendment request: The proposed Technical

Specifications (TS) changes would revise TS Section 2.1 and its

associated TS Basis to reflect the change in the Minimum Critical Power

Ratio (MCPR) Safety Limit due to the plant specific evaluation

performed by General Electric Co. (GE), for LGS Unit 2 Cycle 4.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The revised Minimum Critical Power Ratio (MCPR) Safety Limit for

LGS Unit 2 Technical Specifications, and its use to determine cycle-

specific thermal limits have been performed using NRC-approved

methods within the existing design and licensing basis, and cannot

increase the probability or severity of an accident.

The basis of the MCPR Safety Limit 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 MCPR Safety Limit

preserves the existing margin to transition boiling and fuel damage

in the event of a postulated accident. The probability of fuel

damage is not increased.

Therefore, the proposed TS change does not involve an increase

in the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The MCPR Safety Limit is a Technical Specification numerical

value, designed to ensure that fuel damage from transition boiling

does not occur as a result of the limiting postulated accident. It

cannot create the possibility of any new type of accident. The new

Minimum Critical Power Ratio (MCPR) Safety Limit is calculated using

NRC-approved methods and is based on LGS Unit 2 Cycle 4 specific

inputs.

Therefore, the proposed TS change does not create the

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

accident previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

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

same. The new Minimum Critical Power Ratio (MCPR) Safety Limit is

calculated using NRC approved methods which are in accordance with

the current fuel design and licensing criteria. The MCPR Safety

Limit remains high enough to 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 change does not involve a reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

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

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

County, Alabama

Date of amendments request: September 30, 1996

Description of amendments request: The proposed amendments would

revise Technical Specifications (TSs) 3/4.1.1, 3/4.1.3, 3.1.3.6, 3.2.1,

3/4.2.2, and 3.2.3 and associated Bases to remove certain cycle-

specific parameter limits from the TSs and relocate them to the Core

Operating Limits Report (COLR). These changes result from NRC Generic

Letter (GL) 88-16, dated October 4, 1988, which provided guidance to

licensees on requests for removal of the values of cycle-specific

parameter limits from the TSs. The licensee's proposed amendments are

consistent with the GL.

The COLR has been included in the Definitions section of the TSs.

The definition notes that it is the unit-specific document that

provides these limits for the current operating reload cycle. The

values of these cycle-specific parameter limits are to be determined in

accordance with TS 6.9.1.11. This TS requires that the core operating

limits be determined for each reload cycle in accordance with the

referenced NRC-approved methodology for these limits and consistent

with the applicable limits of the safety analysis. The COLR shall be

provided to the NRC upon issuance.In addition, the above TS changes

would produce administrative changes to the TS Table of Contents.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The removal of cycle-specific core operating limits from the FNP

[Farley Nuclear Plant] Technical Specifications has no influence or

impact on the probability or consequences of a Design Basis Accident

(DBA) occurrence. The cycle-specific core operating limits, although

not in Technical Specifications, will be followed in the operation

of FNP. The proposed amendment retains the same required actions to

be taken when or if limits are exceeded as stipulated by current

Technical Specifications. In addition, the associated surveillance

requirements are not altered by the proposed changes.

Each accident analysis addressed in the FNP FSAR [Final Safety

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

dependent parameters, which are obtained from application of the

NRC-approved reload design methodologies, to ensure that the

transient evaluation of new reloads are bounded by previously

accepted analyses. This examination, which will be performed per

requirements of 10 CFR 50.59, ensures that future reloads will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

As stated earlier, the removal of the cycle-specific variables

has no influence or impact, nor does it contribute in any way to the

probability or consequences of an accident. No safety-related

equipment, safety function, or plant operation will be altered as a

result of this proposed change. The cycle-specific variables are

calculated using the NRC-approved methods and submitted to the NRC

to allow the Staff to continue to trend the values of these limits.

The Technical Specifications will continue to require operation

within the required core operating limits and appropriate actions

will be taken when or if limits are exceeded. Therefore, the

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

kind of accident from any accident previously evaluated.

3. The proposed changes do not result in a significant reduction

in the margin of safety.

The margin of safety is not affected by the removal of cycle-

specific core operating limits from the Technical Specifications.

The margin of safety presently provided by current Technical

Specifications remains unchanged. Appropriate measures exist to

control the values of these cycle-specific limits. The proposed

amendment continues to require operation within the core limits, as

obtained from the NRC-approved reload design methodologies. The

required actions to be taken or if limits are violated remain

unchanged.

The development of the limits for future reloads will continue

to conform to those

[[Page 57492]]

methods described in NRC-approved documentation. In addition, each

future reload involves a 10 CFR 50.59 safety review to assure that

operation of FNP within the cycle-specific limits will not involve a

significant reduction in [the] margin of safety. Therefore, the

proposed changes are administrative in nature and do not impact the

operation of FNP in a manner that involves a reduction to 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: 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

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.

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

NuclearPower Station, Unit No. 1, New London County, Connecticut

Date of amendment request: August 29, 1996

Description of amendment request: The proposed amendment would

modify the applicability requirements for certain radiation monitors so

that the radiation monitors are required to be operable only when

secondary containment integrity is required to be operable; delineate

when secondary containment integrity is required; modify standby gas

treatment operability requirements; make editorial corrections to

clarify the configuration of the radiation monitors; and revise the

associated Bases section.

Date of publication of individual notice in Federal Register:

October 17, 1996 (61 FR 54242)

Expiration date of individual notice: November 18, 1996

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: September 6, 1996

Brief description of amendment request: The proposed amendment

would change Technical Specification (TS) requirements related to steam

generator tubes to allow a laser-welded repair of Westinghouse hybrid

expansion joint (HEJ) sleeved steam generator tubes. Date of individual

notice in Federal Register: October 15, 1996 (61 FR 53769)

Expiration date of individual notice: November 14, 1996

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin

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.

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

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

Units 1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: June 17, 1996

Brief description of amendments: The amendment modifies the

technical specifications (TS) to change (1) the reference method for

calculating dose conversion factors (DCFs) to be used in dose

calculations, and (2) the upper and lower limits for operating

pressurizer pressure to account for new instrument uncertainties and to

reduce the allowed operating band.

Date of issuance: October 23, 1996

Effective date: October 23, 1996, to be implemented within 45 days

of issuance

Amendment Nos.: Unit 1 - 109; Unit 2 - 101; Unit 3 - 81

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 11, 1996 (61

FR 47963). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 23, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

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

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

County, Maryland

Date of application for amendments: June 9, 1995

[[Page 57493]]

Brief description of amendments: The amendments implement changes

to radiological effluent Technical Specifications in accordance with

Generic Letter 89-01 ``Implementation of Programmatic for Radiological

Effluent Technical Specification in the Administrative Controls Section

of the Technical Specifications and Relocation of Procedural Details of

RETS to the Offsite Dose Calculation Manual or to the Process Control

Program.''

Date of issuance: October 18, 1996

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

30 days.

Amendment Nos.: 217 and 194

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 15, 1995 (60 FR

35062) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated October 18, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of application for amendment: May 1, 1996

Brief description of amendment: The proposed amendment will reflect

the implementation of 10 CFR Part 50 Appendix J, Option B at the

Pilgrim Nuclear Power Station.

Date of issuance: October 4, 1996

Effective date: October 4, 1996

Amendment No.: 167

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications.

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

28606) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 4, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Carolina Power & Light Company, et al., Docket No. 50-325,

Brunswick Steam Electric Plant, Unit 1, Brunswick County, North

Carolina

Date of amendment request: April 8, 1996, as supplemented on July

30, 1996, October 4, 1996, October 8, 1996, and October 16, 1996.

Brief description of amendment: The amendment changes the Technical

Specifications to (1) reflect the use of a new type of fuel (GE13) and

(2) modify the minimum critical power ratio safety limit and the

standby liquid control system sodium pentaborate limits to accommodate

the GE13 fuel.

Date of issuance: October 17, 1996

Effective date: October 17, 1996

Amendment No.: 182

Facility Operating License No. DPR-71: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: August 14, 1996 (61 FR

42276) which superseded a Federal Register notice published on June 5,

1996 (61 FR 28607) The Commission's related evaluation of the amendment

is contained in a Safety Evaluation dated October 17, 1996.No

significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan

Date of application for amendment: January 5, 1996, as supplemented

July 12, 1996

Brief description of amendment: The amendment revises the shutdown

cooling (SDC) requirement to allow one train of the SDC system to be

rendered inoperable for testing or maintenance provided that a filled

refueling cavity is available to provide backup decay heat removal

capability in the event that the operating train of SDC becomes

inoperable.

Date of issuance: October 10, 1996

Effective date: October 10, 1996

Amendment No.: 173

Facility Operating License No. DPR-20. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 28, 1996 (61 FR

44348) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 10, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423

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

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

Date of application for amendments: August 8, 1996

Brief description of amendments: The amendments revise the

Technical Specifications, Section 6.9.1.9, to reference updated or

recently approved topical reports used to calculate cycle-specific

limits contained in the Core Operating Limits Report.

Date of issuance: October 24, 1996

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

30 days

Amendment Nos.: 154 and 146

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 11, 1996 (61

FR 47977) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 24, 1996.No significant

hazards consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of amendment request: May 9, 1996

Brief description of amendments: The amendments revised the name

from Arkansas Power & Light Company to Entergy Arkansas, Inc.

Date of issuance: October 23, 1996

Effective date: October 23, 1996

Amendment Nos.: 187 and 177

Facility Operating License Nos. DPR-51 and NPF-6. Amendments

revised the Technical Specifications and the licenses.

Date of initial notice in Federal Register: August 28, 1996 (61 FR

44357) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 23, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi,

Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1,

Claiborne County, Mississippi

Date of application for amendment: May 8, 1996, as supplemented by

letters dated July 18 and September 19, 1996

Brief description of amendment: The amendment modified the

frequency requirements in Surveillance Requirement 3.6.1.3.5 of the

Technical Specifications, on the leakage rate testing for each

containment purge

[[Page 57494]]

isolation valve with resilient seals, to place these purge valves on a

performance basis in accordance with Appendix J of 10 CFR Part 50, as

modified by any exemptions to Appendix J. In addition, the purge valves

would be required to be leakage rate tested every 36 months with at

least two pairs tested every 18 months and, if any purge valve fails to

meet the leakage rate acceptance criterion, all remaining valves must

be tested within 92 days (i.e., a quarter of a year) if not

successfully tested within the previous 92 days.

Date if issuance: October 18, 1996

Effective date: October 18, 1996

Amendment No.: 128

Facility Operating License No. NPF-29. Amendment revises the

Technical Specifications.

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

28614) The additional information contained in the supplemental letters

dated July 18 and September 19, 1996, revised the proposed amendment in

the application of May 8, 1996; however, the revisions were within the

scope of the initial notice and did not affect the staff's proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 18, 1996.No significant hazards

consideration comments received: No

Local Public Document Room location: Judge George W. Armstrong

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

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi,

Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1,

Claiborne County, Mississippi

Date of application for amendment: June 20, 1996, as supplemented

by the letter of September 11, 1996

Brief description of amendment: The amendment redefined the

secondary containment boundary to allow the enclosure building to be

inoperable during the upcoming refueling Outage 8 (RFO 8) scheduled to

begin in October 1996. The amendment added a condition to the license

that the enclosure building may be inoperable during core alterations

and movement of non-recently irradiated fuel (i.e., fuel that has not

occupied part of a critical reactor core for 12 days) during RFO 8 and

the standby gas treatment (SGT) system may be unable to automatically

start or achieve and maintain the required vacuum, provided the

following conditions exist:

a. All dampers communicating between the auxiliary building and the

enclosure building are closed.

b. The access door between the auxiliary building and the enclosure

building is closed, except when the access opening is being used for

entry and exit.

c. The SGT system is blocked from automatic initiation.

d. The SGT system is available for manual initiation or the actions

for Limiting Condition for Operation 3.6.4.3 in the Technical

Specifications for GGNS are complied with.

The non-recently irradiated fuel is spent fuel that has decayed at

least 12 days after the reactor was shut down for refueling.

Date of issuance: October 18, 1996

Effective date: October 18, 1996

Amendment No: 129

Facility Operating License No. NPF-29. Amendment adds a condition

to the license.

Date of initial notice in Federal Register: July 17, 1996 (61 FR

37299) The additional information contained in the supplemental letter

of September 11, 1996, was clarifying in nature and thus, within the

scope of the initial notice and did not affect the staff's proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated October 18, 1996.No significant hazards consideration comments

received: No

Local Public Document Room location: Judge George W. Armstrong

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

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: February 22, 1996, and as

supplemented by letters dated July 22 and September 20, 1996

Brief description of amendment: The amendment revises Clinton Power

Station Technical Specification 3.4.11, ``Reactor Coolant System (RCS)

Pressure and Temperature (P/T) Limits,'' to incorporate specific P/T

limits for the bottom head region of the reactor vessel, separate and

apart from the core beltline region of the reactor vessel.

Date of issuance: October 23, 1996

Effective date: October 23, 1996

Amendment No.: 109

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

Technical Specifications.

Date of initial notice in Federal Register: April 24, 1996 (61 FR

18169) The letters of July 22 and September 20, 1996, provided

clarifying information and did not alter the staff's initial finding

that the proposed changes involve no significant hazards consideration.

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

Safety Evaluation dated October 23, 1996.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. 50-336,

Millstone Nuclear Power Station, Unit No. 2, New London County,

Connecticut

Date of application for amendment: November 21, 1995

Brief description of amendment: The amendment changes Technical

Specification Section 5.2.2, ``Design Pressure and Temperature,'' to

clarify that the reactor containment design temperature is an

equilibrium liner temperature and not the air temperature. The

supporting Technical Specification Bases is updated to reflect the

change and to include the main steam line break accident, in addition

to the loss-of-coolant accident, as the limiting events affecting the

containment temperature and pressure.

Date of issuance: October 21, 1996

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

within 60 days.

Amendment No.: 204

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

Technical Specifications.

Date of initial notice in Federal Register: December 20, 1995 (60

FR 65684) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 21, 1996.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

Omaha Public Power District, Docket No. 50-285, Fort Calhoun

Station, Unit No. 1, Washington County, Nebraska

Date of amendment requests: July 15, 1996, as supplemented by

letters dated September 3, 1996, October 22, 1996, October 23, 1996,

and August 23, 1996

Brief description of amendment: The amendment revises Technical

Specifications (TS) Section 4.3.2 to allow the use of zircaloy or ZIRLO

fuel

[[Page 57495]]

cladding and to use depleted uranium as reactor fuel material. The

amendment also changes TS Section 5.9.5 to add Westinghouse Topical

Reports, WCAP-12610-P-A, ``VANTAGE + Fuel Assembly Report,'' and WCAP-

13027-P, ``Westinghouse ECCS Evaluation Model for Analysis of CE-

NSSS,'' to the list of approved analytical methods for determining the

core operating limits.

Date of issuance: October 25, 1996

Effective date: October 25, 1996

Amendment No.: 178

Facility Operating License No. DPR-40. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 31, 1996 (61 FR

40026) and August 30, 1996 (61 FR 45995). The September 3, 1996,

October 22, 1996, and October 24, 1996, supplemental letters provided

additional clarifying and correcting information and did not change the

initial no significant hazards consideration determination. The

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

Safety Evaluation dated October 25, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

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

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

Obispo County, California

Date of application for amendments: June 7, 1996

Brief description of amendments: The amendments revised the

combined Technical Specifications (TS) for the Diablo Canyon Nuclear

Power Plant, Unit Nos. 1 and 2 by revising Technical Specifications 3/

4.9.14.1, ``Spent Fuel Assembly Storage - Spent Fuel Pool Region 2,''

and TS 3/4.9.14.3, ``Spent Fuel Assembly Storage - Spent Fuel Pool

Region 1,'' to allow storage of fuel assemblies in a checkerboard

pattern in Region 2 of the spent fuel pool (SFP).

Date of issuance: October 25, 1996

Effective date: October 25, 1996, to be implemented within 30 days

from date of issuance.

Amendment Nos.: Unit 1 - 116; Unit 2 - 114

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 25, 1996 (61

FR 50346) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 25, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

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

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

Obispo County, California

Date of application for amendments: December 19, 1995, as

supplemented by letter dated August 8, 1996.

Brief description of amendments: The amendments revised the

combined Technical Specifications (TS) for the Diablo Canyon Power

Plant Unit Nos. 1 and 2 to relocate Technical Specification (TS) 6.5,

``Review and Audit,'' 6.8, ``Procedures and Programs,'' Sections

6.8.1c., 6.8.1d., 6.8.2, and 6.8.3, in accordance with guidance in an

NRC letter dated October 25, 1993, from William T. Russell to the

chairpersons of industry owners groups and the Commission's Final

Policy Statement on TS Improvements for Nuclear Power Reactors on

relocation of TS that do not satisfy the retention criteria. As part of

the relocation of TS 6.8.2, TS 6.1.1 would be revised to require that

proposed tests, experiments, or modifications that affect nuclear

safety be approved by the plant manager or his designee prior to

implementation.

Date of issuance: October 25, 1996

Effective date: October 25, 1996, to be implemented within 90 days

of issuance.

Amendment Nos.: Unit 1 - 117; Unit 2 - 115

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

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

1633) The August 8, 1996, supplemental letter provided additional

clarifying information and did not change the staff's initial no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated October 25, 1996.No significant hazards consideration

comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments: May 20 and 28, 1996, as

supplemented by letter dated July 25, 1996

Brief description of amendments: These amendments, for both units,

add a reference to the ANF-B critical power correlation to Section

6.9.3.2 of the Technical Specifications (TSs); change the values of the

minimum critical power ratio (MCPR) in TS Sections 2.1 and 3.4.1.1.2,

and make appropriate Bases changes. For Unit 1 only, a reference to ABB

licensing methodology report CENPD-300 (for lead use assemblies being

used in the reactor core during the upcoming operating cycle) is added

to Section 6.9.3.2.

Date of issuance: October 11, 1996

Effective date: For both units, as of date of issuance, to be

implemented within 30 days.

Amendment Nos.: 161 and 132

Facility Operating License Nos. NPF-14 and NPF-22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: Unit 2, August 28, 1996

(61 FR 44362); Unit 1, September 4, 1996 (61 FR 47529)The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated October 11, 1996.No significant hazards consideration

comments received: No

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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: March 29, 1996, as supplemented

July 12, 1996, and September 6, 1996.

Brief description of amendment: The proposed amendment would change

the Indian Point 3 Technical Specifications (TSs) relating to minimum

reactor coolant system (RCS) flow and maximum RCS average temperature

to make these parameters consistent with an assumption of 100% helium

release from the boron coating of the integral fuel burnable absorber

rods. The proposed amendment would also add limits associated with

Departure from Nucleate Boiling to the IP3 Technical Specifications

TSs.

Date of issuance: October 22, 1996

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

30 days

[[Page 57496]]

Amendment No.: 170

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

Technical Specifications.

Date of initial notice in Federal Register: July 17, 1996 (61 FR

37301) August 14, 1996 (61 FR 42283)The Commission's related evaluation

of the amendment is contained in a Safety Evaluation dated October 22,

1996.No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

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

San Diego County, California

Date of application for amendments: July 19, 1995, as supplemented

by letters dated December 22, 1995, and March 26, 1996.

Brief description of amendments: These amendments modify Technical

Specification (TS) 3.3.8, ``Containment Purge Isolation Signal

(CPIS),'' and TS 3.3.9, ``Control Room Isolation Signal (CRIS).'' The

revisions are needed to (1) support the upgrading or replacement of

existing radiation monitoring system with state-of-the-art equipment

that will provide for greater operational flexibility and reliability,

and (2) incorporate minor editorial changes to improve clarity of these

TS sections.

Date of issuance: October 8, 1996

Effective date: October 8, 1996, to be implemented within 30 days

of date of issuance

Amendment Nos.: Unit 2 - Amendment No. 132; Unit 3 - Amendment No.

121

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 27, 1995 (60

FR 49948). The December 22, 1995, and March 26, 1996, letters provided

additional clarifying information and did not change the initial no

significant hazards consideration determination.The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated October 8, 1996.No significant hazards consideration

comments received: No.

Temporary Local Public Document Room location: Science Library,

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

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph

M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

Date of amendment request: March 29, 1996, as supplemented by

letters dated June 27, August 29, and September 16, 1996.

Brief description of amendment: The amendment changes Technical

Specification 3/4.4.6, ``Steam Generators'' and associated Bases to

modify the steam generator repair limit 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.

Date of issuance: October 11, 1996

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

30 days

Amendment No.: 115

Facility Operating License No. NPF-8: Amendment revised the

Technical Specifications.

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

25711) The June 27, August 29, and September 16, 1996, letters provided

additional, clarifying information that did not change the scope of the

March 29, 1996, application and the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 11, 1996. No significant hazards

consideration comments received: No.

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

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

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph

M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

Date of amendment request: April 22, 1996, as supplemented by

letters dated May 3, July 15, August 7 and 30, and September 16, 1996

Brief description of amendment: The amendment changes reflect the

implementation of a new F* criterion based on maintaining existing

safety margins for steam generator tube structural integrity concurrent

with allowances for nondestructive examination eddy current

uncertainty.

Date of issuance: October 11, 1996

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

30 days

Amendment No.: 116

Facility Operating License No. NPF-8: Amendment revises the

Technical Specifications.

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

25713) The May 3, July 15, August 7 and 30, and September 16, 1996,

letters provided clarifying information that did not change the scope

of the April 22, 1996, application and the initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 11, 1996. No significant hazards

consideration comments received: No.

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

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

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

Plant, Unit 1, Rhea County, Tennessee

Date of application for amendment: June 29, 1996

Brief description of amendment: The amendment revises Technical

Specification (TS) Section 5.2.2.f to delete the sentence, ``The

Operations Manager shall hold or have held an SRO [Senior Reactor

Operator] license on a similar unit.'' The revision also indicates that

the Operations Superintendent will hold a valid SRO license on this

unit.

Date of issuance: October 15, 1996

Effective date: Octber 15, 1996

Amendment No.: 4

Facility Operating License No. NPF-90: Amendment revises the TS.

Date of initial notice in Federal Register: September 11, 1996 (61

FR 47983)The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 15, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: July 31, 1996 (TXX-96432) as

supplemented by letters dated August 23 and 27 (TXX-96447 and TXX-

96451), and September 19, 1996 (TXX-96469).

Brief description of amendments: The amendments (1) change the

acceptance values for amperes and voltage for the 18 month surveillance

test of the battery chargers; (2) clarify the meaning of the

[[Page 57497]]

term ``associated inverter'' used in the context of energizing 118-Volt

AC Instrument Buses during MODES 1 through 6; and (3) delete the

protection channel and the vital bus ratings for the 118-Volt AC

Instrument Buses identified for MODES 1 through 4.

Date of issuance: October 22, 1996

Effective date: October 22, 1996

Amendment Nos.: Unit 1 - Amendment No. 53; Unit 2 - Amendment No.

39

Facility Operating License Nos. NPF-87 and NPF-89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 28, 1996 (61 FR

44363) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 22, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019.

Dated at Rockville, Maryland, this 30th day of October 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II,Office of Nuclear Reactor

Regulation

[Doc. 96-28372 Filed 11-5-96; 8:45 am]

BILLING CODE 7590-01-F