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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

biweekly notice. Public Law 97-415 revised section 189

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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 August 30, 1996, through September 13, 1996.

The last biweekly notice was published on September 11, 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 October 25, 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 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

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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.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of amendment request: August 2, 1996

Description of amendment request: The proposed amendment would

eliminate from the licenses the requirement to conduct corrosion

testing for the laser welded steam generator sleeves.

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.

This proposed change only involves deleting laboratory testing

requirements designed to demonstrate service life of laser welded

sleeved tubes in the presence of a crevice. Current inspection

requirements ensure that premature degradation is identified and

that tubes containing degraded sleeve joints are plugged.

Operational primary-to-secondary leakage limits ensure that

appropriate action is taken if sleeve degradation results in

leakage. These actions will ensure that offsite dose will be

maintained within a small percentage of 10 CFR 100 limits. Failure

of a sleeve joint is bounded by the Steam Generator Tube Rupture

event evaluated in the [Updated Final Safety Analysis Report] UFSAR.

Therefore, the laboratory testing to determine service life of

sleeved tube joints in the presence of a crevice does not provide

any further useful data. The change does not result in the

installation of any new equipment, and no existing equipment is

modified.

Therefore, this 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.

This proposed change only addresses deleting the laboratory

testing requirements designed to demonstrate service life of sleeved

tubes in the presence of a crevice. Sleeved tubes will continue to

be inspected and plugged in accordance with existing requirements

which are sufficient to ensure detection and repair of degraded

tubes. Premature degradation of tubes is addressed through primary-

to-secondary leakage monitoring and leakage limits. No new equipment

is being installed and no existing equipment is being modified by

this proposed change. Also, no new system configurations will be

introduced as a result of this proposed change. Therefore, no new or

different failure modes are being introduced by deleting the

laboratory testing.

Thus, this 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 a margin of safety.

This proposed change only involves deleting laboratory testing

requirements designed to demonstrate service life of sleeved tubes

in the presence of a crevice. Sleeve integrity will be monitored

during the operating cycle through the current primary-to-secondary

leakage monitoring program. In the event of premature degradation of

a sleeve joint that results in tube leakage, plant shutdown will

occur as required by Technical Specifications and administrative

requirements in accordance with approved plant procedures. Sleeved

tubes will be monitored for degradation in accordance with the

existing inservice inspection requirements which monitors a minimum

20 percent random sleeve sample size. Any tubes with defective

sleeve joints will be plugged as required by Technical

Specifications. Service life of sleeved tubes in the presence of a

crevice, as predicted by laboratory testing, does not affect the

margin of safety of the plant. Therefore, this proposed change does

not involve a significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

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

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

Date of amendment request: July 15, 1996

Description of amendment request: The proposed amendments would

revise Technical Specifications (TS) and associated Bases to relocate

the fire protection program elements from the TS to the Fire Protection

Program. The affected TS sections are 3/4.3.7.9, ``Fire Detection

Instrumentation;'' 3/4.7.5, ``Fire Suppression Systems;'' 3/4.7.6,

``Fire Rated Assemblies;'' and 6.1.C.4,

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``Fire Brigade Staffing.'' In addition, the amendments revise the

Operating License to replace existing fire protection license

conditions with the NRC's standard fire protection license condition.

These changes are made in accordance with the guidance provided in

Generic Letter (GL) 86-10, ``Implementation of Fire Protection

Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements

from Technical Specifications.'' Also, the May 19, 1995, proposed

revision to remove the fire protection requirements from the TS (60 FR

35067) is withdrawn.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

This amendment request does not involve any actual changes to

the fire protection systems at the station. It involves an

administrative change which relocates the control of the Fire

Protection Program from each unit's operating license and technical

specifications to the station Fire Protection Program, as suggested

in Generic Letters 86-10 and 88-12. Therefore, the relocation of

these controls does not affect the assumptions for any of the

accident analysis contained in Chapter 15 of the [Updated Final

Safety Analysis Report] UFSAR.

The Fire Protection Technical Specifications which are to be

relocated to the Fire Protection Program will be controlled by the

proposed fire protection license condition and 10CFR 50.59. These

controls ensure that the requested changes maintain the same level

of control for the Fire Protection Program as that which currently

exists in the Technical Specifications. Therefore, this change is

administrative in nature and does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

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

from any accident previously evaluated because:

This amendment request does not involve any physical changes to

the fire protection systems or reduce the level of control of the

Fire Protection Program. It therefore does not create the

possibility of a new or different type of accident than any

previously described in the UFSAR.

3) Involve a significant reduction in the margin of safety

because:

The same level of control which is currently applied to the Fire

Protection Program by the limiting conditions for operation and the

surveillance requirements of the technical specifications will be

included in the controls applied by the unit licenses and the Fire

Protection Program. Therefore, the margin of safety as defined in

the technical specification bases will not be reduced by this

proposed amendment.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

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: July 26, 1996, and supplemented on

September 3, 1996

Description of amendment request: The proposed amendments would

allow licensee control of the reactor coolant system (RCS) pressure and

temperature (P/T) limits for heatup, cooldown, low temperature

operation and hydrostatic testing. They would also revise the reactor

vessel material surveillance program specimen withdrawal schedule such

that the Unit 2 removal of capsule X is delayed until 19 Effective Full

Power Years (EFPY). This change affects the schedule for withdrawing

surveillance capsules from the reactor vessel for testing to measure

the impact of neutron irradiation of the vessel material and is

required by Section III.B.3 of 10 CFR Part 50, Appendix H, ``Reactor

Vessel Material Surveillance Program Requirements.'' The schedule must

be approved by the Nuclear Regulator Commission (NRC) before

implementation.

Based on input from the Babcock and Wilcox Owners Group Reactor

Vessel Working Group, the data from Zion, Unit 2, capsule X would be

more useful in the overall Master Integrated Reactor Vessel

Surveillance Program (MIRVP) context if irradiated to the ASTM E185-82

maximum of twice the peak End Of Life (EOL) vessel fluence, because

data at higher fluences is needed to characterize irradiation behavior

at the higher EOL fluences characteristic of other non-Commonwealth

Edison MIRVP vessels. For this reason, the licensee is proposing

withdrawing and testing Zion, Unit 2, capsule X at 19 EFPY, which is

currently estimated to occur at refueling outage Z2R18, in the year

2002.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change revises the 10 CFR 50, Appendix H reactor

vessel material specimen withdrawal schedule. Neither the specimens,

nor the process of withdrawal of the specimens, are considered as

initiators for any previously evaluated accident. Further, data at

all fluence levels of current interest based on ASTM E185-82 has

already been obtained from seven Zion Unit 1 and 2 capsules which

have been tested, and the existing evaluations show the reactor

vessel fracture toughness properties to be as expected, and

providing the required safety margin. Extending the time for

withdrawal of the specimen does not adversely affect the pressure

and temperature limit curves for the reactor vessel. Regulatory

Guide 1.99, Rev. 2, was used to prepare the conservative pressure

and temperature limit curves which continue to be requirements.

Additionally, Zion Station participates in the B&W Owners Group

Reactor Vessel Working Group designed to significantly increase the

amount of PWR surveillance data. Under this Working Group, Zion

Station data contributes to the overall understanding of reactor

vessel material irradiation behavior at high EOL fluences, and

obtains the benefit of data from other plants. This program

complements the Zion Station program so that postponement of the

specimen withdrawal will have minimal impact on the understanding of

the irradiation effects on the Zion Station reactor vessel.

Therefore, this change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed revision to the specimen withdrawal schedule does

not change the system operation or design, and therefore, does not

change the response of any required structures, systems or

components in the mitigation of any evaluated accident. As such,

this change does not involve a significant increase in the

consequences of an accident previously evaluated.

The proposed change relocates the RCS P/T, LTOP [low-temperature

overpressure protection] limitations, and supporting information

from the Technical Specifications to Licensee control, specifically

a Pressure Temperature Limits Report (PTLR). Compliance with these

limitations will continue to be required by the Technical

Specifications, however the limitations themselves will be relocated

to a Licensee controlled document. Changes to these limitations will

be controlled by Section 5.6.6 of the Technical Specifications.

Changes to the RCS P/T limits can only be made in accordance with

the approved methodologies listed in the Technical Specifications

which will, in combination with the limitations that continue to be

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imposed by the Technical Specifications, continue to assure the

function of the reactor vessel as a pressure boundary. Revisions to

the LTOP limits can only be made in accordance with the approved

methodologies listed in the Technical Specifications, with any

resulting setpoint changes controlled through a process which

utilizes 10 CFR 50.59. Therefore, this change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration

of the plant (no new or different equipment will be installed). The

proposed revision to the specimen withdrawal schedule does not

change the system operation or design, and therefore, does not

introduce any new failure mechanisms. The proposed specimen

withdrawal schedule continues to provide the required data for

subsequent reactor vessel evaluations, and previous data has

confirmed the confidence in the integrity of the reactor vessel well

beyond the completion of the evaluations following the proposed

withdrawal. Therefore, this revision to the withdrawal schedule does

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

from any accident previously evaluated.

The proposed change does not necessitate a physical alteration

of the plant (no new or different equipment will be installed). The

Technical Specifications will continue to retain requirements to

maintain the RCS within acceptable operational limitations and to

assure operability of the LTOP system. As such, the Technical

Specifications will continue to require compliance with these

limitations. Thus, this change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

of safety?

The proposed change to the specimen withdrawal schedule will not

result in a significant reduction in a margin of safety because it

has no impact on any safety analysis assumptions. Additionally, data

at all fluence levels of current interest based on ASTM E185-82 has

already been obtained with the seven Zion Unit 1 and 2 capsules

which have been tested, and the existing evaluations show the

reactor vessel fracture toughness properties to be as expected, and

providing the required safety margin. The current pressure and

temperature limits are conservative and also provide sufficient

margin to ensure the integrity of the reactor vessel. The proposed

change to the withdrawal schedule does not adversely impact these

curves. Therefore, this change does not involve a significant

reduction in a margin of safety.

The proposed change will not result in a significant reduction

in a margin of safety because it has no impact on any safety

analysis assumptions. Any future changes to the RCS P/T, LTOP

limits, or supporting information must be performed in accordance

with approved NRC methodologies, and compliance with the limitations

relocated to the PTLR will continue to be required by the Technical

Specifications. Additionally, any revision to the LTOP limits which

result in setpoint changes will be controlled through a process

which utilizes 10 CFR 50.59. Therefore, this 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

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

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,

Michigan

Date of amendment request: September 5, 1996 (NRC-96-0075)

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) sections 2.1.2 and 3.4.1.1 to

incorporate cycle-specific safety limit minimum critical power ratios

(SLMCPRs) for the core that will be loaded during the upcoming

refueling outage expected to commence in November 1996.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. 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 SLMCPRs for Fermi 2 for

incorporation into the TS, 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 which result in a more

restrictive value for the SLMCPR. These calculations do not change

the method of operating the plant and have no effect on the

probability of an accident initiating event or transient. The basis

of the MCPR Safety Limit is to ensure that no mechanistic fuel

damage is calculated to occur if the limit is not violated. The new

SLMCPRs preserve the existing margin to transition boiling and 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 changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed change results from analysis of the Cycle 6 core

reload using the same fuel types as previous cycles. These changes

do not involve any new method for operating the facility and do not

involve any facility modifications. No new initiating events or

transients result from these changes. 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 SLMCPRs are 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 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: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226

NRC Project Director: John Hannon

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

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: July 31, 1996, as supplemented by letter

dated September 5, 1996. These letters supersede the application

submitted in letter dated May 9, 1996, which was noticed in the Federal

Register on June 5, 1996 (61 FR 28614).

Description of amendment request: The amendment request would (1)

increase the safety limit minimum critical power ratio (MCPR) for two

loop operation and single loop operation to 1.12 and 1.14,

respectively, and (2) add a General Electric topical report to the list

of documents describing the analytical methods used to determine the

core operating limits. The proposed changes are to Section 2.1.1,

Reactor

[[Page 50343]]

Core Safety Limits, and Section 5.6.5, Core Operating Limits Report

(COLR), respectively, of the Technical Specifications (TSs). This

amendment would go into effect in Operating Cycle 9, at the end of the

upcoming Refueling Outage 8, and the plant will have a mixed core of

Siemens Power Corporation (SPS) 9x9-5 and General Electric (GE) GE11

reload fuel. The licensee also proposed changes to the Bases of the TSs

associated with the above proposed changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

The Minimum Critical Power Ratio (MCPR) safety limit is defined

in the Bases to Technical Specification 2.1.1 as that limit which

``ensures that during normal operation and during Anticipated

Operational Occurrences (AOOs), at least 99.9% of the fuel rods in

the core do not experience transition boiling.'' The MCPR safety

limit is re-evaluated for each reload and, for GGNS [Operating]

Cycle 9, the analyses have concluded that a two-loop MCPR safety

limit of 1.12 based on the application of the generic GE MCPR

methodology is necessary to ensure that this acceptance criterion is

satisfied. For single-loop operation, a MCPR safety limit of 1.14

based on the generic GE MCPR methodology was determined to be

necessary. Core MCPR operating limits are developed to support the

Technical Specification 3.2 requirements and ensure these safety

limits are maintained in the event of the worst-case transient.

Since the MCPR safety limit will be maintained at all times,

operation under the proposed changes will ensure at least 99.9% of

the fuel rods in the core do not experience transition boiling.

Therefore, The Minimum Critical Power Ratio (MCPR) safety limit

change does not affect the probability or consequences of an

accident.

The implementation of GE's GESTAR-II approved methodology has no

effect on the probability or consequences of any accidents

previously evaluated. One exception to GESTAR is that the mis-

oriented and mis-located bundle events will continue to be analyzed

as accidents subject to the acceptance criteria in the current

licensing basis. The design of the GE11 fuel bundles is such that

the bundles are not likely to be mis-oriented or mis-located and the

normal administrative controls will be in effect for assuring proper

orientation and location. Therefore, the probability of a fuel

loading error is not increased. This analysis ensures that

postulated dose releases will not exceed a small fraction (10

percent) of 10 CFR 100 limits.

Therefore, the consequences of accidents previously evaluated

are unchanged.

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

or different kind of accident from any accident previously

evaluated.

The GE11 fuel to be used in [Operating] Cycle 9 is of a design

compatible with fuel present in the core and used in the previous

cycle. Therefore, the GE11 fuel will not create the possibility of a

new or different kind of accident. The proposed changes do not

involve any new modes of operation, any changes to setpoints, or any

plant modifications. They introduce revised MCPR safety limits that

have been proved to be acceptable for Cycle 9 operation. Compliance

with the applicable criterion for incipient boiling transition

continues to be ensured. The proposed MCPR safety limits do not

result in the creation of any new precursors to an accident.

Therefore, the proposed changes do not create the possibility of

a new or different type of accident from any accident previously

evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

The MCPR safety limits have been evaluated to ensure that during

normal operation and during AOOs [abnormal operating occurrences],

at least 99.9% of the fuel rods in the core do not experience

transition boiling. Therefore, the implementation of the proposed

changes in the MCPR safety limit ensure there is no reduction in the

margin of safety.

As with the current SPC methodology, GGNS will implement only

the NRC-approved revisions to GE's GESTAR methodology. This GE

methodology is similar to those SPC reports currently listed in TS

5.6.5 and it will be applied in a similar, conservative fashion. One

exception to GESTAR is that the mis-oriented and mis-located bundle

events will continue to be analyzed as accidents subject to the

acceptance criteria in the current licensing basis. This analysis

ensures that postulated dose releases will not exceed a small

fraction (10 percent) of 10CFR100 [10 CFR Part 100] limits. On this

basis, the implementation of this GE methodology 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: Judge George W. Armstrong

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

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

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

NRC Project Director: William D. Beckner

Gulf States Entergy, Cajun Electric Power Cooperative, and Entergy

Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: August 1, 1996

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to incorporate requirements

for limiting the time that the hydrogen mixing isolation valves on the

drywell are open. The requirements were contained in the old TSs and

with the conversion to the Improved Standard Technical Specifications,

the requirements were inadvertently changed. The proposed action is to

restore requirements to meet the licensing basis for the River Bend

Station. The proposed amendment would also change the time from 7 days

to 31 days to determine the cumulative time the valves are open.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes in this submittal put the requirements that

were in the original Technical Specifications for the Hydrogen

Mixing System back into the current Technical Specifications. The

changes reenstate into the Technical Specifications limitations that

were previously agreed to between River Bend and the Nuclear

Regulatory Commission in the FSAR Safety Evaluation Report for the

Hydrogen Mixing System.

The River Bend SER states in Supplement 2, Section 6.2.4,

``Since the applicant has not demonstrated that these valves are

capable of closing under accident conditions in the drywell, certain

restrictions apply. Technical Specification 3.6.6.2 specifies that

in Operating Modes 1 and 2, the total number of hours used should

not exceed 5 hours/365 days and in Operating Mode 3 the number of

hours should be limited to 90 hours/365 days.'' To date, the

hydrogen mixing isolation valves have not been fully demonstrated to

be capable of closing under accident conditions in the drywell. The

old Standard Technical Specifications (Attachment 2) used at River

Bend reflected this condition. When conversion to ITS was made,

these requirements were dropped but should not have been. In

addition, the requirement to operate the hydrogen mixing system

every 92 days during Modes 1, 2, and 3 was added without

consideration for the requirements in the River Bend Safety

Evaluation Report.

Consequently, for these proposed change, since the requirements

already exist and are being reenstated into the Technical

Specifications, this change is administrative in nature. The

requirements have remained in place through the SER, but were

[[Page 50344]]

inadvertently removed from the Technical Specifications. This change

places the requirements from the SER back into the Technical

Specifications.

In addition, changing the requirement from the old Technical

Specifications for determining the cumulative time that the hydrogen

mixing inlet and outlet valves are open from every 7 days to every

31 days is again administrative in nature, since this only changes

the frequency with which a given requirement is tracked

administratively. It does not change the actual requirement in any

way.

Consequently, since both of these changes are administrative in

nature and only incorporate requirements into the Technical

Specifications that already existed in the RBS FSAR Safety

Evaluation Report, the changes proposed in this amendment request do

not change the probability or consequences of an accident previously

evaluated.

This proposed change does not involve a change to the plant

design or operation. As a result, the proposed change does not

affect any of the parameters or conditions that could contribute to

the initiation of any accidents.

The changes proposed in this amendment request are

administrative in nature and merely add requirements back into the

Technical Specifications that were inadvertently deleted during the

conversion to ITS. Because of the administrative nature of the

proposed changes, it is not possible to create a new or different

kind of accident from any accident previously evaluated.

The proposed changes in this amendment request reenstate

requirements into the Technical specifications that are contained

present in the RBS FSAR Safety Evaluation Report. These requirements

were inadvertently deleted during the conversion to ITS.

Because of the administrative nature of these Technical

Specification changes, there is no change 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: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

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

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

NRC Project Director: William D. Beckner

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

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

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

Matagorda County, Texas

Date of amendment request: August 15, 1996.

Description of amendment request: The proposed amendments would

remove a requirement for performance of a surveillance incorporating a

high toxic gas test signal.

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.

Analyses were performed to evaluated postulated releases of

potentially hazardous chemicals for their impact on Control Room

habitability. The latest revision of these analyses shows that none

of the potentially hazardous chemicals utilized onsite or in the

surrounding 5-mile radius around the South Texas Project pose a

credible hazard to the Control Room. Consequently, there is no need

to ensure that the Control Room Makeup and Cleanup Filtration System

can automatically switch into a recirculation mode of operation by

isolating the normal supply and exhaust flow in response to a High

Toxic Gas test signal. Therefore, elimination of the unnecessary

surveillance has no effect on the probability of an accident or its

consequences.

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

or different kind of accident from any accident previously

evaluated.

The Toxic Gas Monitoring System was provided to protect against

hazardous toxic gas releases only. Verifying automatic switch into

the recirculation mode of operation is no longer necessary since the

Toxic Gas Analyzers have been removed. This change does not affect

other tests for verification of automatic switching into the

recirculation mode of operation. 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 a margin of safety.

Analyses have shown that none of the chemicals onsite and within

a 5-mile radius of the South Texas Project pose a credible hazard to

the facility. Automatic switching of the Control Room Makeup and

Cleanup Filtration System will continue to be verified using test

signals from other sources.

Based upon this evaluation, the South Texas Project has

concluded that these changes do not involve any significant hazards

considerations.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869

NRC Project Director: William D. Beckner

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: August 15, 1996

Description of amendment request: A Federal Register Notice on May

22, 1996 (61 FR 25707), stated that revisions were being proposed to

Clinton Power Station Technical Specification (TS) 3.3.6.2, ``Secondary

Containment Isolation Instrumentation;'' TS 3.3.7.1, ``Control Room

Ventilation System Instrumentation;'' TS 3.6.1.2, ``Primary Containment

Air Locks;'' TS 3.6.1.3, ``Primary Containment Isolation Valves;'' TS

3.6.4.1, ``Secondary Containment;'' TS 3.6.4.2, ``Secondary Containment

Isolation Dampers;'' TS 3.6.4.3, ``Standby Gas Treatment;'' TS 3.7.3,

``Control Room Ventilation;'' and TS 3.7.4, ``Control Room AC System.''

By letter dated August 15, 1996, the licensee revised their proposal to

consolidate the above changes under a newly proposed Special Operations

LCO (i.e., LCO 3.10.10, ``Single Control Rod Withdrawal - Refueling'').

Therefore, the Description of Amendment Request to the TSs has changed

as described herein. The Basis for No Significant Hazards Consideration

has not changed and is repeated below.

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 eliminate CORE ALTERATIONS as an

applicable condition requiring operability of the primary and

secondary containment and control room ventilation system. As stated

in the BASES for the associated Technical Specifications,

operability of these systems is primarily required for mitigation of

the design basis accident - fuel handling accident (DBA-FHA) and

design basis accident - loss of coolant accident (DBA-LOCA). The

performance of CORE ALTERATIONS alone is neither a

[[Page 50345]]

precursor to, nor a condition during which these DBAs are postulated

to occur. The proposed changes only delete CORE ALTERATIONS as an

applicable condition for the affected Technical Specifications. All

other applicable MODES or specified conditions, including operations

with the potential for draining the reactor vessels (OPDRVs) and the

movement of irradiated fuel assemblies within the primary or

secondary containment, remain unchanged. Further, the limitations

placed on the handling of light loads are also unchanged. The

Technical Specifications (and the separate requirements imposed on

the handling of light loads) will thus continue to require that

systems or functions designed to mitigate design-basis/previously

evaluated accidents are OPERABLE during the relevant operating MODES

or conditions. On the basis of the above, it is concluded that the

requested amendment will not increase the probability or

consequences of any accident previously evaluated.

2. The proposed changes do not involve any modification to the

plant design or to the operation of plant systems (except to

determine when certain analyzed accident-mitigating systems or

features are required to be OPERABLE). The failure modes considered

for the proposed changes are the same as those previously

considered, therefore, it can be concluded that no new failure modes

will be created. On this basis, the proposed amendment will not

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

any accident previously evaluated.

3. The changes being made to eliminate CORE ALTERATIONS as an

applicable condition for which certain LCOs must be met, do not

eliminate the requirements for operability of those systems or

features assumed to mitigate design-basis or analyzed accidents

during the applicable MODES when such systems or features are

assumed to be available for performing their mitigating function.

The safety margins assumed or established by the accident analyses

for those design-basis events (as described in the accident analyses

of the Clinton Power Station Updated Final Safety Analysis Report)

therefore remain unchanged. Further, the proposed changes do not

impact the controls imposed on the handling of light loads

(including unirradiated fuel assemblies) for ensuring that such

activities cannot result in an event that yields consequences more

severe than those calculated for the DBA-FHA. With respect to

reactivity concerns during refueling operations (MODE 5), all

systems or features required to be OPERABLE for precluding

inadvertent criticality and monitoring reactivity changes will

continue to be required OPERABLE as per the current Technical

Specification requirements. The deletion of CORE ALTERATIONS as an

applicable condition only applies to the noted systems which do not

contribute to precluding reactivity events. Based on the above, 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

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

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

Atomic Power Station, Lincoln County, Maine

Date of amendment request: August 12, 1996

Description of amendment request: The proposed amendment would add

an additional circumstance to Exception 2 of Technical Specification

(TS) 3.6, Emergency Core Cooling and Containment Spray Systems, during

which operation of a service water/component cooling pump subsystem is

permitted at reduced flow to flush the service water header or inlet

strainer. The Bases for this TS would be augmented to support the

additional circumstance of reduced service water flow.

The proposed amendment would also modify the valve surveillance

requirements of TS 4.6.A.1.b, Periodic Testing of ECCS Valves, to

provide an exception to surveillance requirements for those locked

valves that are inaccessible during power operations or located in a

locked high radiation area. The Bases for this TS would be augmented to

support the change in surveillance 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. The NRC staff's analysis is presented below.

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Invocation of the proposed addition to Exception 2 to TS 3.6 would

not alter any associated Remedial Action completion time, nor those of

TS 3.0.A, Nonconformance with a Limiting Condition for Operation. The

evolutions for which this amendment is intended (flushing a heat

exchanger inlet strainer or cleaning a service water header that has

become fouled)are administratively controlled by procedures that

require review and approval by the Plant Operation Review Committee.

The proposed change to TS 4.6.A.1.b would revise the surveillance

requirements for a very limited number of locked manual valves in the

emergency core cooling system (ECCS). The purpose of the surveillance

requirements is unchanged and is intended to verify that locked valves

remain in their correct position. The position of the valves is not

changed and the revised surveillance requirements will continue to

demonstrate ECCS valve operability.

Thus, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed addition to Exception 2 to TS 3.6 recognizes that

service water cleaning and flushing are operations that are required to

maintain heat transfer capability and equipment reliability. The

proposed amendment does not affect the design of the plant and do not

permit operation of the plant outside the currently allowed modes of

operation.

The proposed change to TS 4.6.A.1.b maintains verification of ECCS

valve operability, while requiring no changes in system configuration

to perform surveillance testing. System functional performance is not

adversely affected.

Thus, the proposed amendment 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 a margin of safety.

The proposed change to TS 3.6 does not significantly alter the

availability or condition of applicable equipment and therefore does

not alter the accident analyses or the conclusions associated with

that equipment. The proposed change permits service water flow to be

reduced below that required for operation of the ECCS in the

recirculation mode, for a short time. The time during which flow is

reduced and both the mussel control and flushing evolutions are

administratively controlled by procedures reviewed and approved by

the Plant Operation Review Committee.

The proposed change to TS 4.6.A.1.b maintains verification of valve

operability. Valve position surveillances will continue to be conducted

in accordance with plant Technical Specifications to ensure valve

operational readiness.

Thus, there is no significant reduction in a 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

[[Page 50346]]

determine that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Wiscasset Public Library, High

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

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

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

John A. Zwolinski

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 amendment requests: June 7, 1996

Description of amendment requests: The proposed amendments would

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

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

(TS) 3/4.9.14.1, ``Spent Fuel Assembly Storage - Spent Fuel Pool Region

2,'' and 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.

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.

Analysis indicates that allowing fuel storage in a checkerboard

pattern with empty storage cells in region 2 of the spent fuel

pool will not result in an inadvertent criticality event. The

keff will continue to remain below 0.95 as required to meet the

acceptance criteria in the NRC Standard Review Plan, Section 9.1.1.

Therefore, the 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 change to allow fuel storage in a checkerboard pattern with

no minimum burnup requirements in region 2 of the spent fuel pool

would designate locations where a fuel assembly could be incorrectly

placed. However, the incorrect placement of a fuel assembly has been

analyzed and would not cause an inadvertent criticality or any other

accident.

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 change does not involve a significant reduction

in a margin of safety.

The NRC Standard Review Plan, Section 9.1.1, acceptance

criterion of a keff of 0.95 provides the margin to criticality.

An analysis was performed that concluded that the proposed change to

allow fuel storage in spent fuel pool region 2 in a checkerboard

pattern meets the acceptance criterion.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

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

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

NRC Project Director: William H. Bateman

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,

Alabama

Date of amendment request: June 6, 1996 (TS 372)

Description of amendment request: The proposed amendment revises

Section 6 of the Browns Ferry Nuclear Plant Units 1, 2, and 3 technical

specifications. Administrative controls associated with quality

assurance are relocated to the licensee's Nuclear Quality Assurance

Plan, consistent with Administrative Letter 95-06, and provides

revisions that make Section 6 more consistent with the improved

Standard Technical Specifications. Additional administrative changes

are included to ensure consistent terminology within the

specifications, and to update obsolete items such as titles and

addresses. The proposed amendment also includes minor editorial

changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed TS change to revise items 1 through 28 above

(Section I, Description of the Proposed Change) was evaluated and

the proposed TS changes were determined to be administrative in

nature. The changes [items 2 through 9, 11, 17 through 21, 23, 26,

and 27] involve administrative title changes of TVA management

positions, the updating of an NRC mailing address and an NRC

regional office title. In addition, certain sections [items 1, 10,

12, 13, 24, and 25] are being relocated into other licensee

documents for which those provisions are adequately controlled by

regulatory requirements. [Items 14, 15, 16, 22, and 28 are editorial

changes.] These changes do not affect any of the design basis

accidents. They do not involve an increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed TS change to revise items 1 through 28 above

(Section I, Description of the Proposed Change) was evaluated and

the proposed TS changes were determined to be administrative in

nature. The changes involve administrative title changes of TVA

management positions, the updating of an NRC mailing address and an

NRC regional office title. In addition, certain sections are being

relocated into other licensee documents for which those provisions

are adequately controlled by regulatory requirements. These changes

do not affect any of the design basis accidents. No modifications to

any plant equipment are involved. There are no effects on system

interactions made by these changes. They do not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed TS change to revise items 1 through 28 above

(Section I, Description of the Proposed Change) was evaluated and

the proposed TS changes were determined to be administrative in

nature. The changes involve administrative title changes of TVA

management positions, the updating of an NRC mailing address and an

NRC regional office title. In addition, certain sections are being

relocated into other licensee documents for which those provisions

are adequately controlled by regulatory requirements. The margin of

safety as reported in the basis for the TSs is not reduced. The

proposed change is administrative and does not impact any technical

information contained in the bases of the TS.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

[[Page 50347]]

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,

Alabama

Date of amendment request: August 30, 1996 (TS 380)

Description of amendment request: The proposed amendment deletes

License Condition 2.C.(3) regarding thermal water quality standards

from the licenses for the Browns Ferry Nuclear Plant Units 1, 2, and 3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed License Condition change is an adminstrative change

and has no relationship to plant safety analyses. Therefore, this

change does not increase the frequency of the precursors to design

basis events or operational transients analyzed in the BFN [Browns

Ferry Nuclear Plant] Final Safety Analysis Report. Likewise, the

proposed changes will not increase the consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed License Condition change is an administrative

change and has no relationship to plant safety analyses. Thus, the

change does not create any type of new accident sequences. Likewise,

the proposed amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed License Condition change is an administrative

change and has no relationship to plant safety analyses. Therefore,

the proposed amendment does not involve a reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

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

NRC Project Director: Frederick J. Hebdon

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

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: August 16, 1996

Description of amendment request: This notice relates to your

submittal to remove the uncertainty term from the specified distance

and remove the footnote which specifies the time frame it is

applicable.

Date of publication of individual notice in Federal Register:

September 11, 1996 (61 FR 47968)

Expiration date of individual notice: October 11, 1996

Local Public Document Room location: location: Waukegan Public

Library, 128 N. County Street, Waukegan, Illinois 60085.

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 3, 1996

Description of amendment request: This notice relates to your

submittal to modify Technical Specification Section 4.3.1.B.4.A.10.a

which provides the acceptance criteria for steam generator tube repairs

by adding a footnote which references the cleanliness and

nondestructive examination requirements as described in CEN-629-P,

Revision 00, ``Repair of Westinghouse Series 44 and 51 Steam Generator

Tubes Using Leak Tight Sleeves.'' Date of publication of individual

notice in Federal Register: September 11, 1996 (61 FR 47966)

Expiration date of individual notice: October 11, 1996

Local Public Document Room location: location: Waukegan Public

Library, 128 N. County Street, Waukegan, Illinois 60085.

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power

Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of amendment request: March 25, 1996, as supplemented by

letter dated August 23, 1996

Brief description of amendment request: The proposed amendment

would revise the safety limit minimum critical power ratios (SLMCPRs)

to support use of GE-13 fuel at PBAPS, Units 2 and 3. Date of

publication of individual notice in Federal Register: August 30, 1996

(61 FR 45997)

Expiration date of individual notice: September 30, 1996

Local Public Document Room location: Government Publications

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

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

PA 17105.

Pennsylvania Power and Light Company, Docket No. 50-387 Susquehanna

Steam Electric Station, Unit 1, Luzerne County, Pennsylvania

Date of amendment request: May 28, 1996, as supplemented by letter

dated July 25, 1996

Brief description of amendment request: The proposed amendment

would revise the Minimum Critical Power Ratio safety limit values,

adding two references to reflect the use of the ANF-B Critical Power

Ratio Correlation and to reflect the use of the ABB Combustion

Engineering licensing methodology, with a modification to the

associated Bases.

Date of publication of individual notice in Federal Register:

September 9, 1996 (61 FR 47529)

Expiration date of individual notice: October 9, 1996

Local Public Document Room location: Osterhout Free Library,

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

[[Page 50348]]

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.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of application for amendment: January 30, 1996, as

supplemented May 20, 1996

Brief description of amendment: This amendment revises the

Technical Specifications (TS) to: (1) add TS 4.6.1.5 to provide

criteria for 24-hour full-load testing of the emergency diesel

generators (EDGs) to be performed during each refueling outage; (2)

revise TS 4.6.1.2 to allow testing of the EDG protective bypasses

listed in TS 3.7.1.d to be done independent of the safety injection or

loss of offsite power testing; and (3) revise TS 4.6.1.3 to include the

EDG protective bypass inspection.

Date of issuance: September 11, 1996

Effective date: September 11, 1996

Amendment No. 174

Facility Operating License No. DPR-23. Amendment revises the

Technical Specifications.

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

FR 7546) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 11, 1996. The May 20,

1996, letter provided clarifying information that did not change the

initial proposed no significant hazards consideration determination. No

significant hazards consideration comments received: No

Local Public Document Room location: location: Hartsville Memorial

Library, 147 West College Avenue, Hartsville, South Carolina 29550

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of application for amendment: December 10, 1995, as

supplemented August 1, 1996, and September 4, 1996.

Brief description of amendment: This amendment revises Technical

Specification (TS) Section 3.5.1 and Tables 3.5-2, 3, and 4 concerning

the reactor trip system, engineering safety feature actuation system,

and isolation function.

Date of issuance: September 12, 1996Effective date: September 12,

1996

Amendment No. 175

Facility Operating License No. DPR-23. Amendment revises the

Technical Specifications.

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

FR 5812). The August 1, 1996, and September 4, 1996, submittals

provided administrative changes to the TS pages that did not change the

initial proposed no significant hazards consideration determination.

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

Safety Evaluation dated September 12, 1996.No significant hazards

consideration comments received: No

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

Station, Unit 1, York County, South Carolina

Date of amendment request: September 30, 1994, as supplemented

September 18, 1995, January 19, March 15, May 16, and August 27, 1996

Description of amendment: The amendment revises the Technical

Specifications to reflect the new setpoints, operational parameters,

and approved analysis methodologies associated with replacement of the

Unit 1 steam generators. The amendment also deletes references to steam

generator tube repair methods, which will no longer be applicable after

the replacement, and clarifies initial surveillances.

Date of issuance: August 29, 1996

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

within 30 days

Amendment No.: 151

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

Technical Specifications.

Date of initial notice in Federal Register: April 10, 1996 (61 FR

15986) The May 16 and August 27, 1996, letters provided clarifying

information that did not change the scope of the September 30, 1994,

application and the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated August 29, 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

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

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: July 17, 1996, as supplemented

August 28, 1996 (TSCR 242, Rev. 2). This application supersedes

applications dated February 23 (TSCR 242) and June 19, 1996 (TSCR 242,

Rev. 1).

Brief description of amendment: The amendment changes the Technical

Specifications (TS) to allow the implementation of 10 CFR Part 50,

Appendix J, Option B.

Date of Issuance: September 3, 1996

Effective date: September 3, 1996, to be implemented within 30 days

of issuance

Amendment No.: 186

Facility Operating License No. DPR-16. Amendment revises the

Technical Specifications.

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

40019) Supersedes notice dated March 27, 1996 (61 FR 13526). The August

28, 1996, supplement provided updated and corrected TS and bases pages.

These

[[Page 50349]]

revisions were within the scope of the original application and did not

change the staff's initial proposed no significant hazards

consideration determination. Therefore renoticing was not warranted.

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

Safety Evaluation dated September 3, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room location: location: Ocean County

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

08753

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, as

supplemented by letter dated July 3, 1996

Brief description of amendment: The amendment revises the Clinton

Power Station Technical Specifications for the drywell to permit bypass

testing on a 10-year frequency with increased testing if performance

degrades, changes the drywell air lock testing and surveillance

requirements, deletes action notes for the drywell air lock and drywell

isolation valves when the bypass leakage limit is not met, and deletes

the specific leakage limits for the drywell air lock seal.

Date of issuance: September 4, 1996

Effective date: September 4, 1996

Amendment No.: 106

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

Technical Specifications.

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

18170) The July 3, 1996, submittal consisted of supporting technical

information which did not change the staff's initial proposed no

significant hazards consideration determination or expand the scope of

the original notice. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated September 4, 1996.

No significant hazards consideration comments received: No

Local Public Document Room location: location: The Vespasian Warner

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

North Atlantic Energy Service Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: June 20, 1996

Description of amendment request: The proposed amendment modifies

the Seabrook Station Appendix A Technical Specifications (TSs) for the

Electrical Power Systems, Onsite Power Distribution. Specifically, the

proposed amendment changes TS 3.8.3.1, Action a., to increase from 8

hours to 7 days the allowable time that 480-volt Emergency Bus

E64 may be less than fully energized.

Date of issuance: August 30, 1996

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment No.: 48

Facility Operating License No. NPF-86. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 26, 1996 (61 FR

33142) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 30, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: location: Exeter Public

Library, Founders Park, Exeter, NH 03833

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

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

Date of application for amendment: April 25, 1996

Brief description of amendment: The amendment modifies the

calibration requirement for the source range monitors and intermediate

range monitors by noting that the sensors are excluded.

Date of issuance: August 19, 1996

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

within 30 days.

Amendment No.: 96

Facility Operating License No. DPR-21. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 19, 1996 (61 FR

31183) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 19, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: 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

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: March 28, 1996

Brief description of amendment: The amendment changes Technical

Specification 3.7.7, ``Sealed Source Contamination,'' and its Bases

that modify the criteria for testing sealed sources for contamination

and leakage. The approved changes are consistent with the testing

criteria currently used at the Millstone Nuclear Power Station, Unit

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

Date of issuance: September 4, 1996

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

within 60 days.

Amendment No.: 202

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

Technical Specifications.

Date of initial notice in Federal Register: May 8, 1996 (61 FR

20853) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 4, 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

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

Power Plant, Unit 3, Humboldt County, California

Date of application for amendment: March 13, 1996

Brief description of amendment: This amendment revised the

Technical Specification by incorporating position changes to reflect a

proposed plant staff reorganization.

Date of issuance: September 6, 1996

Effective date: This license amendment is effective as of the date

of its issuance and must be fully implemented no later than 30 days

from the date of issuance.

Amendment No.: 31Facility License No. DPR-7: This amendment revised

the Technical Specifications

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

18174) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 6, 1996. No

significant hazards consideration comments received: No

Local Public Document Room location: Humboldt County Library, 1313

3rd Street, Eureka, California 95501

[[Page 50350]]

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: February 23, 1996, as

supplemented by letter dated June 28, 1996

Brief description of amendments: These amendments change the

Technical Specification Requirement 4.6.2.1d concerning drywell-to-

suppression chamber bypass testing interval to correspond with the

interval for Primary Containment Integrated Leak Rate Testing under 10

CFR Part 50, Appendix J, Option B.

Date of issuance: September 6, 1996

Effective date: September 6, 1996

Amendment Nos.: 160 and 131

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 10, 1996 (61 FR

15992) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 6, 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

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: June 21, 1996, as supplemented

August 19, 1996, and August 21, 1996.

Brief description of amendment: The amendment extends the

surveillance interval on certain instruments from 18 to 24 months.

Date of issuance: September 5, 1996

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

within 30 days.

Amendment No.: 168

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

Technical Specifications.

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

49027) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 5, 1996. No

significant hazards consideration comments received: No

Local Public Document Room location: location: White Plains Public

Library, 100 Martine Avenue, White Plains, New York 10610.

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: March 6, 1996, as supplemented

by letter dated May 30, 1996.

Brief description of amendment: The amendment changes Technical

Specification (TS) 3.8.1, ``A.C. Sources - Operating,'' to decrease the

minimum fuel oil storage capacity of the Emergency Diesel Generator

Fuel Oil Storage Tanks, from 48,800 to 44,800 gallons. In addition,

footnote ** is deleted from TS 3.8.1.1.b.2. The TS change also adds an

Action Statement to address remedial action when a fuel oil transfer

pump becomes inoperable.

Date of issuance: September 10, 1996

Effective date: As of date of issuance, to be implemented within 90

days.

Amendment No.: 96

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

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

34897) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 10, 1996. No

significant hazards consideration comments received: No

Local Public Document Room location: location: Pennsville Public

Library, 190 S. Broadway, Pennsville, New Jersey 08070

Southern California Edison Company, et al, Docket No. 50-206, San

Onofre Nuclear Generating Station, Unit No. 1, San Diego County,

California

Date of application for amendment: December 22, 1995

Brief description of amendment: The change revises the San Onofre

Unit 1 License Condition to delete a reference to License Condition

2.C(4) from License Condition 2.D. This change eliminates a reporting

requirement for violations of the physical protection plans that is

redundant to reporting requirements in 10 CFR 73.71 and 10 CFR Part 73

Appendix G.

Date of issuance: August 30, 1996

Effective date: August 30, 1996 and shall be implemented no later

than 30 days from August 30, 1996.

Amendment No.: 157

Facility Operating License No. DPR-13: The amendment revised the

Technical Specifications.

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

40028) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 30, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Science Library, University of

California, Irvine, California 92713

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: June 12, 1996

Brief description of amendments: The amendments revise the reactor

core safety limits, Overtemperature delta T (OTDT) and Overpressure

delta T (OPDT) reactor trip setpoints and allowable values, and the

power distribution limits associated with implementation of Relaxed

Axial Offset Control (RAOC) and FQ surveillance. The amendments

also include changes to the Bases associated with these specifications

and surveillances.

Date of issuance: September 3, 1996

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

30 days

Amendment Nos.: 121 and 113

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise

the Technical Specifications.

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

40029) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 3, 1996. No

significant hazards consideration comments received: No

Local Public Document Room location: location: Houston-Love

Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369,

Dothan, Alabama 36302

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: June 20, 1996

Brief description of amendments: The amendments revise the

Technical Specifications to reflect the implementation of 10 CFR Part

50, Appendix J, Option B.

Date of issuance: September 3, 1996

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

30 days

Amendment Nos.: 122 and 114

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise

the Technical Specifications.

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

40030) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 3, 1996. No

significant hazards consideration comments received: No

[[Page 50351]]

Local Public Document Room location: location: Houston-Love

Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369,

Dothan, Alabama 36302

Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear

Plant, Unit 1, Rhea County, Tennessee

Date of application for amendment: July 31, 1996

Brief description of amendment: The amendment revises Technical

Specification 3.6.12 to allow a one-time extension of the 3-month

surveillance requirement for the ice condenser lower inlet doors.

Date of issuance: September 9, 1996

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

within 30 days

Amendment No.: 3

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

Technical Specifications.

Date of initial notice in Federal Register: August 8, 1996 (61 FR

41431) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 9, 1996. No

significant hazards consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: February 19, 1996, as

supplemented on July 3 and August 26, 1996

Brief description of amendment: The amendment revises Kewaunee

Nuclear Power Plant Technical Specification Section 4.2 and its

associated basis by allowing the application of a voltage-based repair

limit for the steam generator tube support plate intersections

experiencing outside diameter stress corrosion cracking. The repair

criteria are based on guidance provided in Generic Letter 95-05,

``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes

affected by Outside Diameter Stress Corrosion Cracking,'' dated August

3, 1995, and on associated industry guidance.

Date of issuance: September 11, 1996

Effective date: September 11, 1996, and is to be implemented within

30 days of the date of issuance.

Amendment No.: 126

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 10, 1996 (61 FR

15999) The July 3 and August 26, 1996, submittals provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

September 11, 1996. No significant hazards consideration comments

received: No.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey County, Kansas

Date of amendment request: October 24, 1995, and superseded by

letter dated May 16, 1996.

Brief description of amendment: The amendment adopts ASTM D3803-

1989 as the laboratory testing standard for charcoal samples from the

charcoal absorbers in the control room filtration system, control

building pressurization system, and the auxiliary/fuel building

emergency exhaust system. The output of the heaters in the control

building pressurization system is reduced from a nominal 15kW to a

nominal 5kW and the acceptance criterion for the testing of the

charcoal absorbers is changed.

Date of issuance: September 4, 1996

Effective date: September 4, 1996, to be implemented within 120

days of the date of issuance.

Amendment No.: 102

Facility Operating License No. NPF-42. The amendment revised the

Technical Specifications.

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

28622) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 4, 1996. No

significant hazards consideration comments received: No.

Local Public Document Room location: locations: Emporia State

University, William Allen White Library, 1200 Commercial Street,

Emporia, Kansas 66801 and Washburn University School of Law Library,

Topeka, Kansas 66621 Dated at Rockville, Maryland, this 18th day of

September 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

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

Regulation

[Doc. 96-24413 Filed 9-24-96; 8:45 am]

BILLING CODE 7590-01-F