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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from July 20, 1996, through August 2, 1996. The

last biweekly notice was published on July 31, 1996 (61 FR 40013).

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

[[Page 42275]]

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 September 13, 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

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

[[Page 42276]]

amendment which is available for public inspection at the Commission's

Public Document Room, the Gelman Building, 2120 L Street, NW.,

Washington, DC, and at the local public document room for the

particular facility involved.

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

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

County, Maryland

Date of amendments request: July 26, 1996

Description of amendments request: The proposed amendment will

revise the appropriate Technical Specifications and their Bases to

permit the electrosleeving repair technique developed by Framatome

Technologies, Inc. to be used at Calvert Cliffs Nuclear Power Plant

(CCNPP). Electrosleeving is a steam generator tube repair method where

an ultra-fine grained nickel is electrochemically deposited on the

inner surface of a tube to form a structural repair of the degraded

tube. The electrodeposition of nickel provides a continuous

metallurgical bond that eliminates all leak paths and macro-crevices.

The electroformed sleeve provides a structural, leak-tight seal,

without deforming or changing the microstructure of the parent tube.

Thus, unlike the conventional welded sleeves, electrosleeving does not

require a post-installation stress relief.

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 would not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The implementation of the proposed steam generator tube

electrosleeving has been reviewed for impact on the current CCNPP

licensing basis.

Since the electrosleeve is designed using the applicable

American Society of Mechanical Engineers (ASME) Boiler and Pressure

Vessel Code as guidance, it meets the objectives of the original

steam generator tubing. The applied stresses and fatigue usage for

the electrosleeve are bounded by the limits established in the ASME

Code. American Society of Mechanical Engineers Code minimum material

property values are used for the structural and plugging limit

analysis. Mechanical testing has shown that the structural strength

of nickel electrosleeves under normal, upset and faulted conditions

provides margin to the acceptance limits. These acceptance limits

bound the most limiting (three times normal operating pressure

differential) burst margin recommended by Regulatory Guide 1.121.

Burst testing of electrosleeved tubes has demonstrated that no

unacceptable levels of primary-to-secondary leakage are expected

during any plant condition.

As in the original tube, the electrosleeve Technical

Specification depth-based plugging limit is determined using the

guidance of Regulatory Guide 1.121 and the pressure stress equation

of Section III of the ASME Code. A bounding tube wall degradation

growth rate per cycle and a nondestructive examination uncertainty

has been assumed for determining the electrosleeve plugging limit.

Evaluation of the proposed electrosleeved tubes indicates no

detrimental effects on the electrosleeve or electrosleeve-tube

assembly from reactor system flow, primary or secondary coolant

chemistries, thermal conditions or transients, or pressure

conditions as may be experienced at Calvert Cliffs. Corrosion

testing of electrosleeve-tube assemblies indicates no evidence of

electrosleeve or tube corrosion considered detrimental under

anticipated service conditions.

The implementation of the proposed electrosleeve has no

significant effect on either the configuration of the plant, or the

manner in which it is operated. The hypothetical consequences of

failure of the electrosleeved tube is bounded by the current steam

generator tube rupture analysis described in Section 14.15 of the

Calvert Cliffs Updated Final Safety Analysis Report. Due to the

slight reduction in diameter caused by the sleeve wall thickness,

primary coolant release rates would be slightly less than assumed

for the steam generator tube rupture analysis (depending on the

break location), and therefore, would result in lower total primary

fluid mass release to the secondary system.

Therefore, BGE [Baltimore Gas and Electric] has concluded that

the proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

of accident from any other accident previously evaluated.

As discussed above, the electrosleeve is designed using the

applicable ASME Code as guidance; therefore, it meets the objectives

of the original steam generator tubing. As a result, the functions

of the steam generators will not be significantly affected by the

installation of the proposed electrosleeve. Adhesion and ductility

tests performed per ASTM [American Society for Testing and

Materials] standards verified that the electrosleeve will not fail

by de-bonding or cracking. In addition, the proposed electrosleeve

does not interact with any other plant systems. Any accident as a

result of potential tube or electrosleeve degradation in the

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

accident analysis. The continued integrity of the installed

electrosleeve is periodically verified by the Technical

Specification requirements.

The implementation of the proposed electrosleeves has no

significant effect on either the configuration of the plant, or the

manner in which it is operated. Therefore, BGE concludes that this

proposed change does not create the possibility of a new or

different kind of accident from any previously evaluated.

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

safety.

The repair of degraded steam generator tubes via the use of the

proposed electrosleeve restores the structural integrity of the

faulted tube under normal operating and postulated accident

conditions. The design safety factors utilized for the electrosleeve

are consistent with the safety factors in the ASME Boiler and

Pressure Vessel Code used in the original steam generator design.

The repair limit for the proposed electrosleeve is consistent with

that established for the steam generator tubes. The portions of the

installed electrosleeve assembly which represent the reactor coolant

pressure boundary can be monitored for the initiation and

progression of electrosleeve/tube wall degradation, thus satisfying

the requirements of Regulatory Guide 1.83. Use of the previously

identified design criteria and design verification testing assures

that the margin to safety with respect to the implementation of the

proposed electrosleeve is not significantly different from the

original steam generator tubes.

Therefore, BGE concludes that the proposed changes does not

involve a significant reduction in a margin of safety.

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

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

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

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

Carolina

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

30, 1996. This notice supersedes the Federal Register notice published

on June 5, 1996 (61 FR 28607).

Description of amendment request: The licensee has proposed to

revise the Technical Specifications (TS) to include the following

changes: 1. The Minimum Critical Power Ratio (MCPR) Safety Limit

specified in TS 2.1.2 from 1.07 to 1.10 for Unit 1 Cycle 11 operation;

TS 5.3.1 to reflect the new fuel type (GE13) that will be inserted

during Unit 1 Refueling Outage 10; 2. The acceptable range of sodium

pentaborate concentration for the standby liquid control system shown

in TS Figure

[[Page 42277]]

3.1.5-1 to reflect changes to poison material concentration needed to

achieve reactor shutdown based on the new GE13 fuel type.

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 license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Proposed Change 1:

The proposed license amendment will allow the loading and use of

GE13 fuel assemblies in the Brunswick Unit 1 reactor core. The use

of GE13 fuel assemblies requires that the safety limit minimum

critical power ratio value also be revised. The safety limit minimum

critical power ratio is established to maintain fuel cladding

integrity during operational transients. The GE13 fuel assembly

design has been analyzed using methods that have been previously

approved by the Nuclear Regulatory Commission and documented in

General Electric Nuclear Energ's reload licensing methodology

Topical Report NEDE-24011, ``General Electric Standard Application

for Reactor Fuel (GESTAR II).``Based on a cycle-specific calculation

performed by General Electric, a safety limit minimum critical power

ratio value of 1.10 has been established for the GE13 fuel type for

Brunswick Unit 1 Cycle 11 operation. The cycle-specific calculation

has been performed in accordance with the methodology in Revision 12

of NEDE-24011. This cycle-specific calculation has demonstrated that

a safety limit minimum critical power ratio value of 1.10 will

ensure that 99.9 percent of the fuel rods avoid boiling transition

during a transient event when all uncertainties are considered. The

safety limit minimum critical power ratio value of 1.10 assures that

fuel cladding protection equivalent to that provided with the

existing safety limit minimum critical power ratio value is

maintained. This ensures that the consequences of previously

evaluated accidents are not significantly increased.

The proposed revision of the safety limit minimum critical power

ratio does not alter any plant safety-related equipment, safety

function, or plant operations that could change the probability of

an accident. The change does not affect the design, materials, or

construction standards applicable to the fuel bundles in a manner

that could change the probability of an accident.

Proposed Change 2:

The standby liquid control system provides a means of reactivity

control that is independent of the normal reactivity control system.

The standby liquid control system must be capable of assuring that

the reactor core can be placed in a subcritical condition at any

time during reactor core life. Technical Specification Figure 3.1.5-

1 specifies the acceptable range of concentrations and volumes for

sodium pentaborate solution used as a neutron absorber (i.e., for

reactivity control). The portion of the sodium pentaborate

concentration range shown in Technical Specification Figure 3.1.5-1

applicable to the lower range of tank volumes is being revised to

increase the required concentration of sodium pentaborate solution.

This change is needed to account for the additional shutdown

reactivity needed based on the planned use of GE13 fuel assemblies

as reload fuel for the Unit 1 reactor core. Since the standby liquid

control system is independent from the normal means of controlling

reactor core reactivity and not used to control core reactivity

during normal plant operations, the proposed revision to the sodium

pentaborate concentration curve for the standby liquid control

system does not alter any plant safety-related equipment, safety

function, or plant operations that could change the probability of

an accident.

The current volume-concentration range of sodium pentaborate

used in the standby liquid control system will achieve a sufficient

concentration of boron in the reactor vessel to ensure reactor

shutdown. Based on the increased reactivity of the new GE13 reload

fuel assemblies, the required sodium pentaborate volume-

concentration range is being revised to ensure sufficient neutron

absorbing solution is available to achieve reactor shutdown;

therefore, the consequences of an accident previously evaluated are

not significantly increased.

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

new or different kind of accident from any accident previously

evaluated.

Proposed Change 1:

The GE13 fuel assembly has been designed and complies with the

acceptance criteria contained in General Electric Nuclear Energy's

standard application for reactor fuel (GESTAR-II), which provides

the latest acceptance criteria for new General Electric fuel

designs. The similarity of the GE13 fuel design to the previously

accepted GE11 fuel design, in conjunction with the increased

critical power capability of the GE13 fuel design, ensure that no

new mode or condition of plant operation is being authorized by the

loading and use of the GE13 fuel type. The proposed revision of the

safety limit minimum critical power ratio from 1.07 to 1.10 does not

modify any plant controls or equipment that will change the plant's

responses to any accident or transient as given in any current

analysis. Therefore, the proposed change to allow the loading and

use of the GE13 fuel type and the revision of the safety limit

minimum critical power ratio value from 1.07 to 1.10 will not create

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

accident previously evaluated.

Proposed Change 2:

As discussed above, the standby liquid control system provides a

means of reactivity control that is independent of the normal

reactivity control system and is capable of assuring that the

reactor core can be placed in a subcritical condition at any time

during reactor core life. The proposed revision to the sodium

pentaborate concentration range does not modify the standby liquid

control system or its controls, does not modify other plant systems

and equipment, and does not permit a new or different mode of plant

operation. As such, the proposed revision to the minimum pentaborate

concentration value does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed license amendment does not involve a significant

reduction in a margin of safety.

Proposed Change 1:

As previously discussed, the GE13 fuel assembly design has been

analyzed using methods that have been previously approved by the

Nuclear Regulatory Commission and documented in General Electric

Nuclear Energy's reload licensing methodology Topical Report NEDE-

24011, ``General Electric Standard Application for Reactor Fuel

(GESTAR II).``The safety limit minimum critical power ratio value is

selected to maintain the fuel cladding integrity safety limit (i.e.,

that 99.9 percent of all fuel rods in the core are expected to avoid

boiling transition during operational transients). Appropriate

operating limit minimum critical power ratio values are established,

based on the safety limit minimum critical power ratio value, to

ensure that the fuel cladding integrity safety limit is maintained.

The operating limit minimum critical power ratio values are

incorporated in the Core Operating limits Report as required by

Technical Specification 6.9.3.1.

Based on the cycle-specific calculation performed by General

Electric, a safety limit minimum critical power ratio value of 1.10

has been established for the GE13 fuel type for Unit 1 Cycle 11

operation. This cycle-specific calculation has been performed based

on the methodology contained in Revision 12 of NEDE-24011-P-A. The

new GE13 safety limit minimum critical power ratio value of 1.10 for

Unit 1 Cycle 11 operation is based on the same fuel cladding

integrity safety limit criteria as that for the GE11 safety limit

minimum critical power ratio (i.e., that 99.9 percent of all fuel

rods in the core are expected to avoid boiling transition during

operational transients); therefore, the proposed change does not

result in a significant reduction in the margin of safety.

Proposed Change 2:

As previously stated, the purpose of the standby liquid control

is to inject a neutron absorbing solution into the reactor in the

event that a sufficient number of control rods cannot be inserted to

maintain subcriticality. Sufficient solution is to be injected such

that the reactor will be brought from maximum rated power conditions

to subcritical over the entire reactor temperature range from

maximum operating to cold shutdown conditions. General Electric

methodology establishes a fuel type dependent standby liquid control

system shutdown margin to account for calculational uncertainties.

General Electric calculations show that an in-vessel concentration

of 660 ppm will provide a standby liquid control system minimum

shutdown margin in excess of the 3.2% delta k value required for the

GE13 fuel. To achieve an in-vessel concentration of 660 ppm, the

acceptable range of standby liquid control system tank

concentrations is being

[[Page 42278]]

revised for the lower range of tank volumes. Thus, the proposed

revision of the standby liquid control system sodium pentaborate

volume-concentration range ensures that there will not be a

significant reduction in the amount of available shutdown margin

and, therefore, not a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

Raleigh, North Carolina 27602

NRC Project Director: Eugene V. Imbro

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

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

Date of amendment request: June 21, 1996

Description of amendment request: The proposed amendments would

extend the surveillance interval for TS 4.7.2.b and 4.7.2.d related to

testing of the Control Room Emergency Filtration System from 18 months

to 24 months. The amendments would also include a one-time extension of

the allowed outage time for the Control Room and Auxiliary Electric

Equipment Room Emergency Filtration System to allow each subsystem to

be inoperable for up to 30 days during modifications to replace the

existing deep bed charcoal absorbers with tray-type units.

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 Technical Specification change does not involve accident

initiators or initial accident assumptions. The Control Room and

Auxiliary Equipment Room Emergency Filtration System (CREFS) trains

A and B are post-accident atmospheric cleanup components that are

designed to limit the radiation exposure to personnel occupying the

Control Room to 5 rem or less whole body during and following all

design basis accident conditions. Therefore, this Technical

Specification change does not increase the probability of occurrence

of an accident previously evaluated.

CREFS trains A and B are utilized to control the onsite dose to

personnel in the Control Room. This Technical Specification change

extends the [Limiting Condition for Operation] LCO duration for

allowing each train to be inoperable one at a time from 7 days to 30

days total for the current surveillance interval. This change is a

one time change to allow for the repair/replacement work associated

with the corroded filter unit charcoal retaining screens in the high

efficiency charcoal adsorber section of each train. The...normal

preventative maintenance and testing [will] be performed on the

operable CREFS train just prior to taking the [opposite] filter

train out of service for the modification. This action will ensure

that the remaining subsystem is operable and ensure maximum

reliability of the system. The Technical Specification change will

not affect onsite dose if a [design-basis accident] DBA occurs and

the operating filter unit does not fail. The operable filter unit

will be sufficient to maintain the operating areas habitable. The

original LCO allowed 7 day operation with only one operable train

and is also susceptible to a single failure during the Allowed

Outage Time. The probability that a DBA will occur coupled with the

single failure of the operable train during the extended allowed

outage time per the Technical Specification change is the same order

of magnitude as for the current 7 day allowed outage time.

Therefore, this change does not increase the consequences of an

accident previously evaluated.

The extension of the surveillance interval from 18 months to 24

months extends the maximum interval between TS surveillances of the

filter trains from 22.5 months to 30 months. The equipment that is

affected are the CREFS filter trains A and B, which are comprised of

HEPA filters, heaters, charcoal adsorbers, and fans. This equipment

has a history of satisfactory surveillance testing (in-place testing

and laboratory analysis of charcoal), and has had little maintenance

problems for the past 5 years. Although the SER Section 6.4.1 and

the [Regulatory Guide] RG 1.52 state that the units shall be tested

every 18 months, a review of the basis documents for the testing

(ANSI N510) shows that the 1975 edition recommended annual testing

and later editions (1980 and 1989) state that testing be performed

``at least once every operating cycle''. Therefore the extension of

the surveillance intervals from 18 months to 24 months will not

increase the consequences of an accident previously evaluated.

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

from any accident previously evaluated because:

This Technical Specification change will allow each train of

CREFS to be inoperable one at a time for up to 30 days to repair/

replace charcoal retaining screens and changes surveillance

intervals from 18 months to 24 months. Prior to the extended LCO on

a given train, the scheduled monthly surveillance and preventive

maintenance will be performed. This Technical Specification change

does not involve components that are accident initiators and

therefore will not create a new or different kind of accident than

those previously analyzed.

3) Involve a significant reduction in the margin of safety

because:

The purpose of CREFS trains A and B are to control the onsite

dose to personnel in the Control Room following an accident that

involves a potential radiological release. Redundant filter trains

are utilized to ensure that a single active failure will not impact

the ability of the system to perform its safety function. Since the

probability of an accident occurring during the extended Technical

Specification LCO for the inoperable train in conjunction with the

probability that the operable CREFS train will fail is the same

order of magnitude as for the current LCO, then the proposed

Technical Specification change has minimal impact on the safe

operation of the plant. The CREFS trains were both determined

operable following their last surveillance and no events have

occurred at the plant to indicate that they may be inoperable.

Normal preventative maintenance and testing will be performed on the

operable CREFS train just prior to taking the [opposite] filter

train out of service for the modification. This action will ensure

that the remaining subsystem is operable and ensure maximum

reliability of the system. The change in surveillance intervals from

18 months to 24 months will not cause a significant reduction in the

margin of safety, because the previous five surveillances have been

satisfactory and the equipment/components do not have a tendency to

drift over time. Therefore, the proposed amendment will not

significantly impact the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: 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

Dairyland Power Cooperative (DPC), Docket No. 50-409, LaCrosse

Boiling Water Reactor (LACBWR), Vernon County, Wisconsin

Date of amendment request: April 10, 1996

Description of amendment request: The proposed amendment would

update the facility Possession Only License and Technical

Specifications to reflect the permanently shutdown and defueled

condition of the plant.

[[Page 42279]]

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:

DPC proposes to modify the LACBWR Technical Specifications to

more accurately reflect the permanently shutdown, defueled,

possession-only status of the facility.

Analysis of no significant hazards consideration:

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

the probability or consequences of an accident previously evaluated.

The proposed changes delete system requirements that are no

longer necessary to prevent, or mitigate the consequences of, a

credible SAFSTOR accident as described in our current SAFSTOR

Accident Analysis.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes are either administrative in nature or were

made based on the analysis of previously evaluated accident

scenarios. In no other way do they change the design or operation of

the facility and therefore do not create the possibility of a new or

different kind of accident from any previously evaluated.

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

in the margin of safety.

The changes incorporate into the proposed Technical

Specifications the margin of safety associated with the current

SAFSTOR accident analysis and thus don't involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: LaCrosse Public Library, 800

Main Street, LaCrosse, Wisconsin 54601.

Attorney for licensee: Wheeler, Van Sickle and Anderson, Suite 801,

25 West Main Street, Madison, Wisconsin 53703-3398

NRC Project Director: Seymour H. Weiss

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

Michigan

Date of amendment request: July 25, 1996 (NRC-96-0064)

Description of amendment request: The proposed amendment would

relocate or delete a number of items currently in the Administrative

Controls Section (Section 6.0) of the technical specifications (TS).

This submittal revises a previous submittal dated December 15, 1994

(NRC-94-0107), to modify the proposed TS change to be consistent with

NRC Administrative Letter 95-06, ``Relocation of Technical

Specifications Administrative Controls Related to Quality Assurance,''

the Improved Standard TS (ISTS), and pending changes to the ISTS. The

previous submittal was noticed in the Federal Register on June 6, 1995

(60 FR 29873).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated

because the proposed changes are administrative in nature. None of

the proposed changes involve a physical modification to the plant, a

new mode of operation or a change to the UFSAR [Updated Final Safety

Analysis Report] transient analyses. No Limiting Condition for

Operation, ACTION statement or Surveillance Requirement is affected

by any of the proposed changes.

Also, these proposed changes, in themselves, do not reduce the

level of qualification or training such that personnel requirements

would be decreased. Therefore, this change is administrative in

nature and does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Further, the proposed changes do not alter the design, function, or

operation of any plant component and therefore, do not affect the

consequences of any previously evaluated accident.

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

or different kind of accident from any accident previously evaluated

because the proposed changes do not introduce a new mode of plant

operation, surveillance requirement or involve a physical

modification to the plant. The proposed changes are administrative

in nature. The changes propose to revise, delete or relocate the

stated administrative control provisions from the TS to the UFSAR,

plant procedures or the QA [Quality Assurance] Program whereby,

adequate control of information is maintained. Further, as stated

above, the proposed changes do not alter the design, function, or

operation of any plant components and therefore, no new accident

scenarios are created.

3. The proposed changes do not involve a significant reduction

in a margin of safety because they are administrative in nature.

None of the proposed changes involve a physical modification to the

plant, a new mode of operation or a change to the UFSAR transient

analyses. No Limiting Condition for Operation, ACTION statement or

Surveillance Requirement is affected. The proposed changes do not

involve a significant reduction in a margin of safety. Additionally,

the proposed change does not alter the scope of equipment currently

required to be OPERABLE or subject to surveillance testing nor does

the proposed change affect any instrument setpoints or equipment

safety functions. Therefore, the change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: Mark Reinhart

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

Unit No. 1, Pope County, Arkansas

Date of amendment request: April 29, 1996

Description of amendment request: The proposed amendment revises

the permissible values of the maximum and minimum pressurizer water

levels and incorporates a graph to display these values for various

operating conditions. The amendment also revises the Bases section of

the Technical Specification. The Bases changes revise the acceptable

value of the as-found tolerance for the settings of the pressurizer

safety valves and change the value of flowrate through the pressurizer

safety valves. The moderator temperature coefficient as described in

the Bases Section is removed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

Consequences of an Accident Previously Evaluated.

The startup accident and the rod withdrawal accident have been

reanalyzed to justify the proposed increase in pressurizer coder

safety value as-found tolerance. The analyses establish more

appropriate boundaries and re-analyze the same initiators as are

currently found in the ANO-1 Safety Analysis Report. Changing the

as-found setpoint tolerance does not change how the pressurizer code

safety valve operates as it will continue to be reset to 2500 psig

plus or minus 1% prior to reactor startup.

The acceptance criteria for these analyses are that the reactor

coolant system (RCS)

[[Page 42280]]

pressure shall not exceed the safety limit of 2750 psig (110% of

design pressure and that the reactor thermal power remains below

112% Rated Power. The analyses using the proposed setpoint tolerance

have shown that the acceptance criteria were met and that the

consequences of the events were essentially the same as those in the

ANO-1 SAR. Analyses were performed to determine the pressurizer

maximum water level that would prevent the RCS from exceeding the

safety limit of 2750 psig in the event of either a startup accident

or a rod withdrawal accident. More appropriate pressurizer level

requirements have been incorporated in accordance with these

analyses.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

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

Accident from any Previously Evaluated.

The proposed changes introduce no new mode of plant operation.

The reanalysis of the startup accident and the rod withdrawal

accident were performed using methodologies identical to that

employed in the ANO-1 SAR and an improved computer code (RELAP5/

MOD2). The pressurizer code safety valve setpoint will continue to

be reset at 2500 psig plus or minus 1% prior to reactor startup and

will continue to function to maintain RCS pressure below the safety

limit of 2750 psig. Analyses were performed to determine the

pressurizer maximum water level that would prevent the RCS from

exceeding the safety limit of 2750 psig in the event of either a

startup accident or a rod withdrawal accident. More appropriate

pressurizer level requirements have been incorporated in accordance

with these analyses.

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

or different kind of accident from any previously evaluated.

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

Safety.

The safety function of the pressurizer code safety valves is not

altered as a result of the proposed change in setpoint tolerance.

The reanalysis of the startup accident and rod withdrawal accident

have shown that with a plus or minus 3% setpoint tolerance, the

pressurizer code safety valves will function to limit RCS pressure

below the safety limit of 2750 psig. The sensitivity studies for the

startup accident showed the acceptance criteria would still be met

even if one pressurizer code safety valve lifted at 5% above 2500

psig at startup conditions. Additional analyses were performed to

determine the pressurizer maximum water level that would prevent the

RCS from exceeding the safety limit of 2750 psig in the event of

either a startup accident or a rod withdrawal accident.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

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

NRC Project Director: William D. Beckner

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

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

Date of amendment request: June 28, 1996

Description of amendment request: The proposed amendments would

remove the Unit 1 and Unit 2 Technical Specification requirements to

secure the containment equipment hatch during core alterations or fuel

handling.

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:

Criterion 1 - Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The proposed change would allow the containment equipment hatch

door to remain open during fuel movement and core alterations. This

door is normally closed during this time period in order to prevent

the escape of radioactive material in the event of a fuel handling

accident. This door is not an initiator of any accident. The

probability of a fuel handling accident is unaffected by the

position of the containment equipment hatch door. The current fuel

handling analysis, which has been approved by the Staff for ANO-2

and submitted for ANO-1, calculates maximum offsite doses to be well

within the limits of 10 CFR Part 100. The current fuel handling

accident analysis results in maximum offsite doses of 63.6 and 41.8

Rem to the Thyroid and 0.902 and 0.598 Rem to the whole body (sum of

beta and gamma) for ANO-1 and ANO-2, respectively. This analysis

assumes the entire release from the damaged fuel is allowed to

migrate to the site boundary unobstructed. Therefore, allowing the

equipment hatch doors to remain open results in no change in

consequences. Also, the calculated doses during a fuel handling

accident would be considerably larger than the actual doses since

the calculation does not incorporate the closing of the equipment

hatch door following evacuation of containment. The proposed change

would significantly reduce the dose to workers in the containment in

the event of a fuel handling accident by expediting the containment

evacuation process. Therefore, this change does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

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

Different Kind of Accident from any Previously Evaluated.

The proposed change does not involve the addition or

modification of any plant equipment. Also, the proposed change would

not alter the design, configuration, or method of operation of the

plant beyond the standard functional capabilities of the equipment.

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

different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

The proposed change does not have the potential for an increased

dose at the site boundary due to a fuel handling accident. The

margin of safety as defined by 10 CFR Part 100 has not been

significantly reduced. Closing the equipment hatch door following an

evacuation of containment further reduces the offsite doses in the

event of a fuel handling accident and provides additional margin to

the calculated offsite doses. Therefore, this change does not

involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

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

NRC Project Director: William D. Beckner

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

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

Date of amendment request: July 12, 1996

Description of amendment request: The proposed amendment would

change Technical Specification (TS) Sections 6.2.2.h and 6.2.2.i. To

provide adequate shift coverage without routine heavy use of overtime,

TS Section 6.2.2.h specifies an objective to have operating personnel

work ``a normal 8-hour day, 40-hour week'' while the facility is

operating. The proposed amendment would change the objective to ``an 8

to 12 hour day, nominal 40-hour week.''

TS Section 6.2.2.i currently states, ``The General Supervisor

Operations, Supervisor Operations, Station Shift Supervisor Nuclear,

and Assistant Station Shift Supervisor Nuclear shall hold senior

reactor operator licenses.'' The proposed amendment would change this

section to state, ``The

[[Page 42281]]

Manager Operations, Station Shift Supervisor Nuclear and Assistant

Station Shift Supervisor Nuclear shall hold senior reactor operator

licenses.'' This change is based upon a reorganization that eliminates

the positions of General Supervisor Operations and Supervisor

Operations from the Unit 1 Operations management structure. The

responsibilities of these positions will be assumed by the Manager

Operations or delegated to off-shift Senior Reactor Operators. Thus,

Senior Reactor Operators will report directly to the Manager

Operations.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

proposed amendment, will not involve a significant increase in the

probability or consequence of an accident previously evaluated.

Establishing operating personnel work hours at, ``an 8 to 12

hour day, nominal 40-hour week,'' provides enhanced continuity for

normal plant operations. There has been no noticeable increase in

safety related problems during the trial period [The facility has

been implementing 12-hour operator shifts for over 1 year on a trial

basis]. Overtime remains controlled by site administrative

procedures in accordance with the NRC Policy Statement of working

hours (Generic Letter 82-12). The probability for operating

personnel error due to (1) incomplete or insufficient turnover or

(2) interruption of in-plant maintenance and testing is reduced. No

physical plant modifications are involved, and none of the

precursors of previously evaluated accidents are affected.

Therefore, this change will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The assimilation of the responsibilities of the previous

positions of General Supervisor Operations and Supervisor Operations

into the position of Manager Operations and to off-shift Senior

Reactor Operators reflects a restructuring of the operations

department, and is essentially a reduction in layers of management.

This proposed change does not involve any physical modification to

the plant, and does not affect any precursor of a previously

evaluated accident. Therefore, this change will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

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

different kind of accident from any accident previously evaluated.

Establishing operating personnel hours at ``an 8 to 12-hour day,

nominal 40-hour week'' provides increased flexibility in scheduling

and does not adversely affect their performance. Overtime remains

controlled by site administrative procedures in accordance with the

NRC Policy Statement on working hours (Generic Letter 82-12). No

physical modification of the plant is involved. As such, the change

does not introduce any new failure modes or conditions that may

create a new or different accident. Therefore, operation in

accordance with the proposed amendment will not create the

possibility of a new or different kind of accident from any

previously evaluated.

The responsibilities of the previous positions of General

Supervisor Operations and Supervisor Operations will be assimilated

into the positions of the Manager Operations and the off-shift

Senior Reactor Operators. There is no physical plant modification.

The change does not introduce any new failure modes or conditions

that may create a new or different accident. Therefore, the change

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

of accident from any accident previously evaluated.

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

proposed amendment, will not involve a significant reduction in a

margin of safety.

Establishing operating personnel hours at ``an 8 to 12-hour day,

nominal 40-hour week,'' provides increased flexibility in scheduling

and does not adversely affect their performance. This change also

decreases the risk of miscommunication between shifts by reducing

the number of turnovers per day and increases operations and

maintenance efficiency by promoting continuity in ongoing plant

activities. Overtime remains controlled by site administrative

procedures in accordance with the NRC Policy Statement on working

hours (Generic Letter 82-12) and is consistent with the Improved

Standard Technical Specifications. The proposed change involves no

physical modification of the plant, or alterations to any accident

or transient analysis [...], and the changes are administrative in

nature. Therefore, the change does not involve any significant

reduction in a margin of safety.

The assimilation of the responsibilities of the positions of

General Supervisor Operations and Supervisor Operations, into the

positions of the Manager Operations and the off-shift Senior Reactor

Operators, effectively reduces layers of management. The proposed

change is consistent with Standard Review Plan (SRP) 13.1.2-13.1.3.

This administrative transformation of the operations department

management structure involves no physical modification of the plant

or alterations to any accident or transient analysis. Therefore,

this change in itself does not involve any significant reduction in

a margin of safety.

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

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

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

amendment request

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

Point Nuclear Station Unit No. 2, Oswego County, New York

Date of amendment request: July 12, 1996

Description of amendment request: The proposed amendment would

change Technical Specification (TS) Section 6.2.2.i. To provide

adequate shift coverage without routine heavy use of overtime, TS

Section 6.2.2.i specifies an objective to have operating personnel work

``a normal 8-hour day, 40-hour week'' while the facility is operating.

The proposed amendment would change the objective to ``an 8 to 12 hour

day, nominal 40-hour week.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

proposed amendment, will not involve a significant increase in the

probability or consequence of an accident previously evaluated.

Establishing operating personnel work hours at, ``an 8 to 12

hour day, nominal 40-hour week,'' allows normal plant operations to

be managed more effectively and with enhanced continuity. There has

been no noticeable increase in safety related problems during the

trial period [The facility has been implementing 12-hour operator

shifts for over 1 year on a trial basis]. Overtime remains

controlled by site administrative procedures in accordance with the

NRC Policy Statement on working hours (Generic Letter 82-12). The

probability for operating personnel error due to (1) incomplete or

insufficient turnover or (2) interruption of in-plant maintenance

and testing is reduced. No physical plant modifications are

involved, and none of the precursors of previously evaluated

accidents are affected. Therefore, this change will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

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

different kind of accident from any accident previously evaluated.

Establishing operating personnel hours at, ``an 8 to 12-hour

day, nominal 40-hour week,'' improves the quality of life for

operating personnel and does not adversely affect their performance.

Overtime remains controlled by site administrative procedures in

accordance with the NRC Policy Statement on working hours (Generic

Letter 82-12). No physical modification of the plant is

[[Page 42282]]

involved. As such, the change does not introduce any new failure

modes or conditions that may create a new or different accident.

Therefore, operation in accordance with the proposed amendment will

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

from any previously evaluated.

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

proposed amendment, will not involve a significant reduction in a

margin of safety.

Establishing operating personnel hours at, ``an 8 to 12-hour

day, nominal 40-hour week,'' improves the quality of life for

operating personnel and does not adversely affect their performance.

This change also decreases the risk of miscommunication between

shifts and increases operations and maintenance efficiency by

promoting continuity in ongoing plant activities. Overtime remains

controlled by site administrative procedures in accordance with the

NRC Policy Statement on working hours (Generic Letter 82-12) and is

consistent with the Improved Standard Technical Specifications. The

proposed change involves no physical modification of the plant, or

alterations to any accident or transient analysis [...], and the

changes are administrative in nature. Therefore, the change does not

involve any significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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 amendment request: February 2, 1996

Description of amendment request: This request would change

Technical Specification (TS) 3.6.1.2 for each unit to permit primary

containment leakage testing of the main steam isolation valves (MSIVs)

at either 22.5 psig or 45 psig according to the type of test to be

conducted. Currently the TS only specifies 22.5 psig for the MSIVs'

test pressure.

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. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change to the allowable test pressure for MSIV leak

testing was reviewed from two perspectives. First is the potential

for the change in testing pressure, and test methodology, to impact

testing results. The second perspective is the potential for a

failure of the testing configuration to result in undesirable

consequences.

Under the proposed change, an increased test pressure of 45.0

psig (Pa) in the accident direction will be used to perform

Technical Specification required MSIV leak testing. However, the

acceptance criteria for testing is maintained consistent with

current Technical Specifications. Therefore, the proposed change to

allow a test pressure of Pa will not affect the validity of

leak test results. The existing Technical Specification required

leak integrity of the MSIVs will be maintained under the proposed

test methodology and thus the ability of the MSIVs to act as a

containment isolation valves is not affected.

The proposed test pressure of Pa will be applied in the

accident direction, and will result in a back pressure being applied

to the Main Steam Line (MSL) Plugs. The potential for MSL Plug

ejection has been reviewed and adequate precautions have been taken

to ensure that fuel damage would not result from [local leak rate

test] LLRT induced MSL Plug ejection. The MSL Plugs are installed

using a restraint ring which prevents inadvertent ejection.

[Pennsylvania Power and Light Company] PP&L procedures require that

the restraint ring be installed as a prerequisite for LLRT testing

of the MSIVs at Pa. However, in the unlikely event that the MSL

Plug and restraint ring were installed improperly and then subjected

to back pressurization at Pa, ejection could occur. If this

event did occur, the MSL Plug could hit the fuel which is an

accident bounded by the fuel assembly handling accident analysis

addressed in [Final Safety Analysis Report] FSAR Section 15.7.4. The

MSL Plugs, MSL Plug Restraint Ring, and MSL Plug Insert and Remove

Tool meet the requirements of NUREG 0612 and PP&L's Heavy Loads

Program.

Therefore, the proposal to allow an alternative test pressure,

Pa, does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

II. This proposal does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

All components within the test volume have been evaluated for

structural integrity under the proposed test pressures. In addition,

pressurization of the Main Steam Line Plugs during testing will be

below the evaluated pressure. The acceptance criteria for the test

will be maintained, thus verification of the leak integrity of the

MSIVs will not be impacted. Therefore, the proposed change to allow

for an alternative test pressure of (Pa) does not create the

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

previously evaluated.

III. This change does not involve a significant reduction in a

margin of safety.

The proposed change does not affect the acceptance criteria for

the MSIV LLRT. As a result, testing at Pa in the accident

direction will provide an equivalent test to that which is performed

at Pa. No change in the leak integrity of the MSIVs is

anticipated as a result of performing the testing at the alternative

pressure. The potential for MSL Plug ejection during MSIV LLRT at

Pa has been evaluated and found to be bounded by existing

accident analysis. Therefore the proposed change to allow an

alternative test pressure, Pa, 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: Osterhout Free Library,

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

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: John F. Stolz

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

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

Date of amendment request: July 12, 1996

Description of amendment request: The proposed amendment would

revise the Indian Point 3 (IP3) Technical Specifications (TSs) by

changing the surveillance frequency requirements in Table 4.1-1,

``Minimum Frequencies for Checks, Calibrations, and Tests of Instrument

Channels'' to accommodate a 24-month operating cycle.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response:

The proposed changes do not involve a significant increase in

the probability or consequence of any accident previously evaluated.

The proposed changes are being made to extend surveillance

frequencies from 18 months to 24 months for:

[[Page 42283]]

Vapor Containment High Radiation Monitors

Reactor Coolant System Subcooling Margin Monitor (SMM),

Overpressure Protection System (OPS), and

Reactor Vessel Level Indication System (RVLIS).

These proposed changes are being made using the guidance

provided by Generic Letter 91-04 to accommodate a 24-month fuel

cycle. The containment radiation monitors, SMM, and RVLIS are used

to provide operator information during post-accident conditions and

have no effect on event initiators associated with previously

analyzed accidents. The OPS is used only when the plant is shutdown,

with RCS [reactor coolant system] temperature below a low

temperature limit, and the RCS is not vented. The function of the

OPS is to protect the RCS from Low Temperature Overpressurization

(LTOP) transients and has no effect on accident initiators. No

credit is taken in the IP3 safety analyses for accident mitigation

effects that might result from use of these instrument channels.

Updated calculations and evaluations to assess the proposed increase

in the surveillance intervals demonstrate that the effectiveness of

these instrument channels in fulfilling their respective functions

is not reduced. The containment high radiation monitors are used for

post accident monitoring purposes to provide operators with an

indication of adverse conditions in containment based on releases of

radioactivity from the RCS to the containment atmosphere. These

monitors provide no signals to plant control systems or automatic

safety systems used for accident mitigation and have no role as an

accident initiator.

Use of the subcooling margin monitor and core exit thermocouples

by plant operators is specified in the Indian Point 3 Emergency

Operating Procedures (EOPs) to assess post accident cooling

conditions in the RCS. Changes to the EOPs will be made to reflect

the results of the updated loop accuracy calculations for this

instrumentation. These changes will ensure that safety analysis

input assumptions associated with subcooling margin, for small break

LOCA [loss-of-coolant accident], steam generator tube rupture, and

steamline break, remain valid, and that the response strategies

outlined in the Westinghouse Owners Group Emergency Response

Guidelines are maintained. Core exit thermocouple readings are not

used for input to plant safety analyses.

The OPS provides a protective function to prevent RCS pressure

limits from being exceeded while the plant is shutdown and the RCS

is being maintained at a low temperature and not vented. Failure of

the OPS is not assumed to be an accident initiator in the plant

safety analyses.

The change to the RVLIS calibration interval does not affect

design or operation of plant systems and will not affect the

probability of accidents. Revised loop accuracy calculations have

demonstrated that operator actions for responding to postulated

accidents using RVLIS in conjunction with the Indian Point 3 EOPs

will remain consistent with the accuracy requirements RVLIS. The

consequences of a previously evaluated accident will not be

affected.

Equipment and system design requirements and safety analysis

acceptance criteria continue to be met with the proposed new

surveillance intervals. Based on the above information it is

concluded that the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously analyzed.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

The proposed changes to extend the surveillance frequencies for

the above listed instrument channel do not create the possibility of

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

The increased surveillance frequencies were evaluated based on past

equipment performance and do not require any plant hardware changes

or changes in system operation. There are no new failure modes

introduced as a result of extending these surveillance intervals,

which could lead to the creation of new or different kinds of

accident.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed changes do not involve a significant reduction in a

margin of safety. [A decreased] surveillance frequency for the

Containment High Radiation Monitor, SMM, OPS, and RVLIS does not

adversely affect the performance of safety-related systems,

equipment, or instruments and does not result in increased severity

of accidents evaluated. The radiation monitor, SMM, and RVLIS are

not used to support margins of safety identified in the Technical

Specifications. OPS provides an equipment protection function to

prevent inadvertent overpressurization of the RCS at shutdown

conditions. The Low Temperature Overpressurization (LTOP) curve in

the Technical Specifications represents material stress limits based

on fracture toughness requirements for ferritic steel. Analysis of

the proposed change to the OPS surveillance frequency verified

sufficient margin to the LTOP curve and therefore does not involve a

significant reduction in margin to the material stress limits.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

New York, New York 10019.

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

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

Date of amendment request: July 12, 1996

Description of amendment request: The proposed amendment would

change the Indian Point 3 (IP3) Technical Specifications (TS) relating

to minimum reactor coolant system (RCS) flow and maximum RCS average

temperature to make these parameters consistent with an assumption of

100% helium release from the boron coating of the integral fuel

burnable absorber (IFBA) rods.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of any accident

previously evaluated?

The proposed changes to the RCS minimum flow and maximum

Tavg requirements will not increase the probability or

consequences of an accident previously evaluated. Reference 2 [SECL-

96-046, ``IFBA Helium Release Evaluation for Cycle 9 Restart,''

Westinghouse Electric Corporation, dated July 8, 1996] states that,

for the remainder of Cycle 9, all pertinent licensing basis

acceptance criteria have been met, and the margin of safety as

defined in the Technical Specification Bases is not reduced in any

of the licensing basis accident analyses for the assumption of a

100% helium release from the IFBA rods. Reference 3 [Westinghouse

letter, ``Technical Specification Value for T-Average,'' INT-96-557,

dated July 3, 1996] states that a reduction of maximum allowable

indicated Tavg from 578.3 deg.F to 571.5 deg.F specifications

consistent with the more limiting containment integrity analyses.

The associated plant and technical specification changes do not

affect any of the mechanisms postulated in the FSAR [Final Safety

Analysis Report] to cause licensing basis events. Therefore, the

probability of an accident previously evaluated has not increased.

Because design limitations continue to be met, and the integrity of

the RCS pressure boundary is not challenged, the assumptions

employed in the calculation of the offsite radiological doses remain

valid. Therefore, the consequences of an accident previously

evaluated will not be increased.

(2) Does the proposed license amendment create the possibility

of a new or different kind of accident from any previously

evaluated?

The proposed changes to the RCS minimum flow and maximum

Tavg requirements do not create the possibility of a new or

different kind of accident from any previously evaluated. Reference

2 states that, for the remainder of Cycle 9, all pertinent

[[Page 42284]]

licensing basis acceptance criteria have been met, and the margin of

safety as defined in the Technical Specification Bases is not

reduced in any of the licensing basis accident analyses for the

assumption of a 100% helium release from the IFBA. Reference 3

provides clarifications of the assumptions made in the design basis

and restricts DNB temperature limits to be consistent with non-DNB

analyses. The associated plant and technical specification changes

do not change the plant configuration in a way which introduces a

new potential hazard to the plant (i.e., no new failure mode has

been created). Therefore, an accident which is different than any

previously evaluated will not be created.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

The proposed changes to the RCS minimum flow and maximum

Tavg requirements do not involve a significant reduction in a

margin of safety. Reference 2 demonstrates that, for the remainder

of Cycle 9, all pertinent licensing basis acceptance criteria have

been met, and the margin of safety as defined in the Technical

Specification Bases is not reduced in any of the licensing basis

accident analyses for the assumption of a 100% helium release from

the IFBA. Reference 3 maintains the margin of safety by restricting

a DNB limit to bound other analyses. Since References 2 and 3

demonstrate that all applicable acceptance criteria continue to be

met, the subject operating conditions will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

New York, New York 10019.

NRC Project Director: Jocelyn A. Mitchell, Acting

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: May 3, 1996 (TS 352)

Description of amendment request: The proposed amendment requests

administrative changes to the Browns Ferry Nuclear Plant (BFN) Units 1,

2, and 3 technical specifications. The proposed amendment consists of

three parts, designated by the licensee as A, B, and C. Part A deletes

technical specification requirements associated with BFN Unit 2

Amendment 219, issued November 12, 1993, to permit modification of

reactor vessel water level instrumentation requested by NRC Bulletin

93-03. Part B deletes technical specification requirements associated

with Amendment 228, issued on December 7, 1994, which provided a

temporary change to permit upgrade of electrical equipment. The

modifications associated with Parts A and C are complete. Part C

provides other administrative changes to clarify requirements and to

implement rule 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.

Part A: The proposed Technical Specification change to remove

the temporary revisions, which were in place to modify the reactor

vessel water level instrumentation requested by NRC Bulletin 93-03,

is administrative. The temporary limiting condition for the minimum

number of trip systems operable will no longer be accurate and the

minimum number operable per trip system will be the same as they

were prior to November 12, 1993. Therefore, the proposed changes

will not significantly increase the consequences of an accident

previously evaluated.

Part B: The proposed Technical Specification change to remove

the temporary revisions, which were in place to replace the 250 volt

shutdown board batteries is administrative. The LCO to extend the

allowed outage time (AOT) from a five-day to a 45-day AOT will no

longer be accurate and the five day AOT will be the same as it was

prior to Unit 2, Cycle 7. Therefore, the proposed changes will not

significantly increase the consequences of an accident previously

evaluated.

Part C: The proposed Technical Specifications change revises

items 1 through 5 above (Section I, Description of the Proposed

Change, Part C), and is administrative. TVA has evaluated the

proposed technical specification changes and has determined that the

proposed changes are administrative in nature. Further, it provides

a revision based on an NRC Code of Federal Regulations rule change.

Also, the proposed changes provide correction of administrative

errors from previous technical specifications. For example, the Main

Steamline High Radiation remarks in Table 3.2.A, 1.b., should have

been deleted from the TS as part of TS-322. It also clarifies some

requirements to ensure consistent application throughout the

specifications. 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.

Part A: The proposed Technical Specification change to remove

the temporary revisions, which were in place to modify the reactor

vessel water level instrumentation requested by NRC Bulletin 93-03,

is administrative. The temporary limiting condition for the minimum

number of trip systems operable will no longer be accurate and the

minimum number operable per trip system will be the same as they

were prior to November 12, 1993. 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.

Part B: The proposed Technical Specification change to remove

the temporary revisions, which were in place to replace the 250 volt

shutdown board batteries is administrative. The LCO to extend the

allowed outage time (AOT) from a five day to a 45-day AOT will no

longer be accurate and the five day AOT will be the same as it was

prior to Unit 2, Cycle 7. 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.

Part C: The proposed Technical Specifications change revises

items 1 through 5 above (Section I, Description of the Proposed

Change, Part C), and is administrative. TVA has evaluated the

proposed changes and has determined that they are administrative in

nature. Further, it provides revisions based on an NRC Code of

Federal Regulations rule change. It also provides correction of

administrative errors in previous technical specification changes.

For example, the Main Steamline High Radiation remarks in Table

3.2.A, 1.b., should have been deleted from the TS as part of TS-322.

It also clarifies some requirements to ensure consistent application

throughout the specifications. 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 change is administrative in nature for Parts A, B,

and C. The proposed change includes the deletion of temporary

changes as a result of modifications to systems and clarification of

some requirements to ensure consistent application throughout the

specifications. Further, the proposed change corrects errors in

previous TS submittals. No safety margins are affected by these

changes.

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

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

satisfied. Therefore, the NRC staff

[[Page 42285]]

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 l0H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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 l0H, 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: June 21, 1996 (TS 377)

Description of amendment request: The proposed amendment provides a

new minimum critical power ratio safety limit to replace the current

non-conservative value. The amendment also updates the technical

specification bases to clarify the usage of the residual heat removal

supplemental spent fuel pool cooling mode.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change in the Safety Limit Minimum Critical Power

Ratio (SLMCPR) does not increase the frequency of the precursors to

design basis events or operational transients analyzed in the Browns

Ferry Final Safety Analysis Report. Therefore, the probability of an

accident previously evaluated is not significantly increased.

The proposed change in the SLMCPR ensures that 99.9 percent of

the fuel rods in the core are expected to avoid boiling transition

during the most limiting anticipated operational occurrence, which

is the design and licensing basis for the analysis of accidents and

transients described in the Browns Ferry Updated Final Safety

Analysis Report (UFSAR). It does not change the nuclear safety

characteristics of any safety system or containment system.

Therefore, the consequences of an accident, operator error, or

malfunction of equipment important to safety previously evaluated in

the UFSAR has not been increased.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change to the Technical Specification requirements

for the safety limit minimum critical power ratio does not involve a

modification to plant equipment. No new failure modes are

introduced. There is no effect on the function of any plant system

and no new system interactions are introduced by this change.

Therefore, the proposed amendment does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change will ensure that during any anticipated

operational transient, at least 99.9% of the fuel rods would be

expected to avoid boiling transition which is consistent with the

licensing basis. Since the margin [of] safety is being increased

with this change, the proposed amendment 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: 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

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

Callaway County, Missouri

Date of application request: July 18, 1996

Description of amendment request: The amendment adopts ASTM D-3803-

1989 as the laboratory testing standard for charcoal samples from the

charcoal adsorbers in the auxiliary/fuel building emergency exhaust

system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The requested change to the charcoal sample surveillance

acceptance criteria for the fuel building and auxiliary building

emergency exhaust system will not affect the method of operation of

the system. The testing of the charcoal filter samples will continue

to be performed in accordance with NRC-accepted methods and

acceptance criteria, and the new test protocol will still ensure

filter efficiency is maintained equal to or greater than 90%. There

are no changes to the emergency exhaust system and it will continue

to function in a manner consistent with the safety analysis

assumptions and the plant design basis. There will be no degradation

in the performance of or an increase in the number of challenges to

equipment assumed to function during an accident. Therefore, the

proposed changes will not increase the probability or consequences

of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The changes to the surveillance requirements are being made to

adopt current NRC-accepted methods of testing charcoal samples.

These changes will not affect the method of operation of the

applicable systems and the laboratory testing will continue to

demonstrate the required adsorber performance after a design-basis

LOCA [loss-of-coolant accident] or fuel handling accident. No new or

different kind of accident from any previously evaluated will be

created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The new charcoal adsorber sample laboratory testing protocol is

more stringent than the current testing practice and meets current

NRC-approved test methods. The new testing criteria will continue to

demonstrate the required adsorber performance after a design-basis

LOCA or fuel handling accident and will not affect the filter system

performance. Therefore, this change will not reduce the margin of

safety of the emergency exhaust system filter operation.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

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

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

NRC Project Director: William H. Bateman

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: July 18, 1996

Description of amendment request: The proposed amendment would

revise

[[Page 42286]]

Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 3.8,

``Refueling Operations,'' and its associated Basis, by allowing the

containment personnel air lock doors to remain open during refueling

operations as long as at least one door is capable of being closed in

30 minutes or less.

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 were reviewed in accordance with the

provisions of 10 CFR 50.92 to determine that no significant hazards

exist. The proposed changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Maintaining the doors of the personnel air lock open during

REFUELING OPERATIONS does not adversely affect the probability or

consequences of accidents previously evaluated. The only applicable

accident is a fuel handling accident described in [Updated Safety

Analysis Report] USAR Section 14.2.1. The fuel handling accident

evaluated in the USAR Section 14.2.1 assumes the accident to be in

the spent fuel pool in the Auxiliary Building. The accident assumes

a sudden release of the gaseous fission products held in the voids

between the pellets and cladding of all of the rods in the highest

rated fuel assembly at 100 hours following reactor shutdown. The

accident activity is assumed to discharge from the spent fuel pool

directly to the atmosphere at ground level. No credit is taken for

existing building structures, ventilation, or filtration systems. A

fuel handling accident in containment is bounded by this evaluation.

Furthermore, any release from a fuel handling accident in

containment can still be terminated by closing one of the personnel

air lock doors following containment evacuation.

The containment personnel air lock doors are components integral

to the containment structure. They are not accident initiators.

Therefore, the proposed amendment does not increase the probability

of any previously evaluated accident.

The control room operator immersion and inhalation doses were

reviewed as part of the updated Control Habitability Evaluation

Report. The report states that thyroid and whole body doses received

by control room operators in each of the other design basis

accidents discussed in KNPP USAR Section 14.2 are less than the

[loss of coolant accident] LOCA dose. This amendment does not change

the results of the Control Room Habitability Evaluation Report,

since the fuel handling accident evaluated in KNPP USAR Section

14.2.1 assumes a release directly to the atmosphere. This change

does not significantly increase the consequences of an accident

previously evaluated.

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

from any accident previously evaluated.

The accident evaluated in USAR section 14.2.1 bounds a fuel

handling accident in containment with the personnel air lock doors

open. The fuel handling accident evaluated in USAR section 14.2.1

assumes activity is discharged directly to the atmosphere at ground

level. Since no credit is taken for building structures, ventilation

systems or filtration systems, the position of the doors does not

affect the analysis of record. Furthermore, one of the air lock

doors can still be closed following containment evacuation to

terminate the release.

The containment personnel air lock doors are components integral

to the containment structure. They are not accident initiators. The

proposed amendment does not create the possibility of any new or

different kind of accident [from any accident] previously evaluated.

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

Maintaining the containment personnel air lock doors open during

REFUELING OPERATIONS does not involve a significant reduction in the

margin of safety. A fuel handling accident in containment is bounded

by a fuel handling accident in the spent fuel pool. The spent fuel

pool fuel handling accident is assumed to have a sudden release of

the gaseous fission products held in the voids between the pellets

and cladding of all of the rods in the highest rated fuel assembly,

100 hours following reactor shutdown. The accident activity leaving

the spent fuel pool is assumed to discharge directly to the

atmosphere at ground level. No credit is taken for existing building

structures, ventilation, and filtration systems. Therefore, there is

no reduction in the current margin of safety. Furthermore, the

release caused by a fuel handling accident in containment can be

terminated by closing one of the personnel air lock doors following

containment evacuation.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Wisconsin,

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

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P. O. Box 1497, Madison, Wisconsin 53701-1497

NRC Project Director: Gail H. Marcus

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.

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

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: July 12, 1996

Brief description of amendment request: The amendment would change

Technical Specification 3.3.2.1, ``Engineered Safety Feature Actuation

System Instrumentation,'' to reflect a revised setpoint for the

interlock designated P-12.

Date of publication of individual notice in Federal Register: July

23, 1996 (61 FR 38229)

Expiration date of individual notice: August 22, 1996

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

West Broadway, Salem, NJ 08079

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

[[Page 42287]]

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 29, 1996, as

supplemented June 17, 1996.

Brief description of amendment: The amendment revises the technical

specifications (TS) table 4.1-3, item 4 to change the frequency of main

steam safety valve (MSSV) testing to that specified in NUREG-1431, the

improved ``Standard Technical Specifications, Westinghouse Plants'' and

adds the MSSV test acceptance requirements.

Date of issuance: August 1, 1996

Effective date: August 1, 1996

Amendment No.: 171

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

Technical Specifications.

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

FR 7545). The June 17, 1996, submittal provided supplemental

information that was not outside the scope of the February 28, 1996,

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

contained in a Safety Evaluation dated August 1, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

North Carolina

Date of application for amendment: March 20, 1996

Brief description of amendment: To relocate Technical Specification

3.3.3.2, Movable Incore Detectors, to plant procedures.

Date of issuance: July 24, 1996

Effective date: July 24, 1996

Amendment No.: 65

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

Technical Specifications

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

18164) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No

Local Public Document Room location: Cameron Village Regional

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

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County and Northeast Nuclear Energy Company,

et al., Docket Nos. 50-245, 50-336, and 50-423, Millstone Nuclear

Power Station, Units 1, 2, and 3, New London County, Connecticut

Date of application for amendments: November 22, 1995

Brief description of amendments: The amendments replace the title-

specific designation of members representing specific functional areas

on the Plant Operating Review Committee (PORC) for the Haddam Neck

Plant and Millstone Units 1, 2, and 3 with a functional area-specific

designation that stipulates membership qualification and experience

requirements. The amendments also clarify the composition of the Site

Operations Review Committee (SORC) at Millstone.

Date of issuance: July 16, 1996

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

within 60 days.

Amendment Nos.: 190, 95, 200, 130

Facility Operating License Nos. DPR-61, DPR-21, DPR-65, AND NPF-49:

Amendments revised the Technical Specifications.

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

FR 7549) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 16, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Russell Library, 123 Broad

Street Middletown, Connecticut 06457, for the Haddam Neck Plant, and

the Learning Resources Center, Three Rivers Community-Technical

College, 574 New London Turnpike, Norwich, Connecticut 06360, and

Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385, for Millstone 1, 2, and 3.

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

Station, Unit 1, York County, South Carolina

Date of application for amendment: January 26, 1996, as

supplemented May 6, May 20, and June 5, 1996

Brief description of amendment: The amendment revises the Technical

Specifications to permit a one-time operation of the containment purge

ventilation system during Mode 3 and 4 after the steam generator

replacement outage.

Date of issuance: July 30, 1996

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

30 days

Amendment No.: 150

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

Technical Specifications.

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

18165) The supplemental submittals provided clarifying information that

did not change the scope of the January 26, 1996, application for

amendment nor the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated July 30, 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

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

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

Date of application for amendments: March 4, 1996

Brief description of amendments: The amendments delete Flow

Monitoring System from Technical Specification 3.4.6.1 and associated

surveillance requirements.

Date of issuance: July 29, 1996

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

30 days

Amendment Nos.: 168 and 150

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

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

18166) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 29, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

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

[[Page 42288]]

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

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

Date of application for amendments: March 4, 1996

Brief description of amendments: The amendments consist of changes

to the Final Safety Analysis Report for McGuire Units 1 and 2 to delete

the seismic qualification requirement for the Containment Atmosphere

Particulate Radiation Monitors.

Date of issuance: July 30, 1996

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

30 days

Amendment Nos.: 169 and 151

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Final Safety Analysis Report.

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

20845) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 30, 1996, and an

Environmental Assessment dated July 22, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

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

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

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

Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: May 20, 1996

Brief description of amendment: The amendment revised the Facility

Operating License and Appendix C to the license to reflect the name

change from Gulf States Utilities Company to Entergy Gulf States, Inc.

Date of issuance: July 30, 1996

Effective date: July 30, 1996

Amendment No.: 88

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

operating license and Appendix C to the license.

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 July 30, 1996. No significant

hazards consideration comments received. No

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

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

Mississippi Electric Power Association, and Entergy Mississippi,

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

Claiborne County, Mississippi

Date of application for amendment: November 20, 1995, as

supplemented by letter dated December 15, 1995

Brief description of amendment: The amendment revised and deleted

surveillance requirements, notes, and action statements involved with

the requirements for the drywell leak rate testing, and the air lock

leakage and interlock testing in Subsections 3.6.5.1 (Drywell), 3.6.5.2

(Drywell Air Lock), and 3.6.5.3 (Drywell Isolation Valves) of the

technical specifications.

Date of issuance: August 1, 1996

Effective date: August 1, 1996

Amendment No: 126

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

Technical Specifications.

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

25704) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 1, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Judge George W. Armstrong

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

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: March 21, 1996 as supplemented

May 13, 1996.

Brief description of amendments: Relocate requirements for

Radiological Effluent Controls from Technical Specifications (TS) to

the Offsite Dose Calculation Manual or the Process Control Program. New

programmatic controls for radioactive effluent and radiological

environmental controls will be incorporated into the TS. Also,

requirements for Gas Decay tanks and Explosive Gas Mixture will be

placed in a different area of the TS.

Date of issuance: July 31, 1996

Effective date: July 31, 1996

Amendment Nos.: 188 and 182Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 19, 1966 (61 FR

31180) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 31, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: May 28, 1996

Brief description of amendments: Amendment changes Technical

Specification 6.2.2.i, ``Administrative Controls,'' regarding

Operations Manager qualifications.

Date of issuance: July 22, 1996

Effective date: July 22, 1996

Amendment Nos.: 187 and 181Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

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

31181) The Commission's related evaluation of the amendments is

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

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

GPU Nuclear Corporation and Saxton Nuclear Experimental (SNEC)

Corporation, Docket No. 50-146, Saxton Nuclear Experimental

Facility (SNEF)

Date of application for amendment: February 2, 1996, as

supplemented on February 28, April 24, and May 24, 1996.

Brief description of amendment: The proposed amendment would (1)

increase the scope of work permitted at SNEF to include asbestos

removal, removal of defunct plant electrical services, and installation

of decommissioning support facilities and systems; (2) eliminate areas

within the containment vessel requiring administrative access controls;

and (3) revise the facility layout diagram to allow the exclusion area

to consist of, at a minimum, the containment vessel and, at a maximum,

to extend to the SNEF outer security fence and to include on the

diagram the footprint of the proposed decommissioning support

facilities.

Date of issuance: July 23, 1996

Effective date: July 23, 1996

Amendment No.: 14

Amended Facility License No. DPR-4: Amendment changed the Technical

Specifications.

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

31182).

[[Page 42289]]

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

safety evaluation dated July 23, 1996. No significant hazards

consideration comments received: No

Local Public Document Room location: Saxton Community Library, 911

Church Street, Saxton, Pennsylvania 16678

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: February 1, 1996

Brief description of amendment: The amendment revised Technical

Specifications to allow an increase in the initial nominal Uranium-235

enrichment limit for fuel assemblies which may be stored in the spent

fuel pool.

Date of issuance: July 30, 1996

Effective date: July 30, 1996

Amendment No.: 174

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

Technical Specifications.

Date of initial notice in Federal Register: March 13, 1996 (61 FR

10396) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No.

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

South 15th Street, Omaha, Nebraska 68102

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

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

Obispo County, California

Date of application for amendments: May 9, 1996

Brief description of amendments: The amendments revised the

combined Technical Specifications (TS) for the Diablo Canyon Nuclear

Power Plant (DCPP), Unit Nos. 1 and 2 by revising Technical

Specifications (TS) 3/4.3.2, ``Engineered Safety Features Actuation

System Instrumentation,'' and 3/4.6.2, ``Containment Spray System.''

The changes clarified the description of the initiation signal required

for operation of the containment spray system at DCPP and correctly

incorporated changes made in previous license amendments. All of the

changes are administrative in nature.

Date of issuance: August 1, 1996

Effective date: August 1, 1996

Amendment Nos.: Unit 1 - 114; Unit 2 - 112

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

revised the Technical Specifications.

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

31184) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 1, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

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

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

San Diego County, California

Date of application for amendments: June 3, 1996, as superseded by

application dated June 25, 1996.

Brief description of amendments: These amendments revise Improved

Technical Specification (TS) 3.3.11, ``Post Accident Monitoring

Instrumentation (PAMI),'' and Improved TS 5.5.2.13, ``Diesel Fuel Oil

Testing Program.'' Specifically, the number of instruments required to

measure reactor coolant inlet temperature (TCold), and reactor

coolant outlet temperature (THot), will be revised from two per

loop to two (with one cold leg indication and one hot leg indication

per steam generator). These changes to the Improved TS reinstate

provisions of the current San Onofre Nuclear Generating Station

(SONGS), Unit Nos. 2 and 3 TS revised as part of NRC Amendment Nos. 127

and 116 for SONGS Units 2 and 3 (referred to as the Improved TS).

Date of issuance: August 1, 1996

Effective date: August 1, 1996, to be implemented by August 9,

1996.

Amendment Nos.: Unit 2 - 130; Unit 3 - 119

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 2, 1996 (61 FR

34452) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 1, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

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

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and

50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa

County, Virginia

Date of application for amendments: July 26, 1995, as supplemented

April 25, 1996. The April 25, 1996, letter provided clarifying

information that did not change the scope of the July 26, 1995,

application and initial proposed no significant hazards consideration

determination.

Brief description of amendments: The amendments clarify the

Technical Specifications to allow switching of charging and low-head

safety injection pumps during unit shutdown conditions. These

amendments also allow additional methods of rendering these same pumps

incapable of injecting into the reactor coolant system when required

for low-temperature conditions.

Date of issuance: July 24, 1996

Effective date: July 24, 1996

Amendment Nos.: 202 and 183

Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: August 30, 1995 (60 FR

45190) The Commission's related evaluation of the amendments is

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

hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: May 8, 1996

Brief description of amendment: The amendment revises Kewaunee

Nuclear Power Plant Technical Specification (TS) 5.3, ``Reactor,'' and

TS 5.4, ``Fuel Storage,'' by removing the enrichment limit for reload

fuel and imposing fuel storage restrictions on the spent fuel storage

racks and the new fuel storage racks. The revised TS are structured

consistent with the Westinghouse Standard Technical Specifications and

the fuel storage restrictions are based on the criticality analyses

used to support Amendment No. 92 dated March 7, 1991.

Date of issuance: July 23, 1996

Effective date: July 23, 1996

Amendment No.: 124

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

Technical Specifications.

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

31185) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No.

[[Page 42290]]

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: May 1, 1995

Brief description of amendment: This amendment revises TS Section

6.0, throughout, to reflect an organization change in which the

position of Vice President Plant Operations has been eliminated and the

positions of Chief Operating Officer and Plant Manager were created.

This change assigns certain management responsibilities to the Chief

Operating Officer and Plant Manager.

Date of issuance: August 1, 1996

Effective date: August 1, 1996, to be implemented within 30 days of

issuance.

Amendment No.: 100

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

Technical Specifications.

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

25716) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 1, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room 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 7th day of August 1966.

For the Nuclear Regulatory Commission

Steven A. Varga, Director,

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

Regulation

[Doc. 96-20586 Filed 8-13-96; 8:45 am]

BILLING CODE 7590-01-F