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

UNITED STATES 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 September 30, 1996, through October 10,

1996. The last biweekly notice was published on October 9, 1996 (61 FR

52962).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By November 22, 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

[[Page 55029]]

controverted. In addition, the petitioner shall provide a brief

explanation of the bases of the contention and a concise statement of

the alleged facts or expert opinion which support the contention and on

which the petitioner intends to rely in proving the contention at the

hearing. The petitioner must also provide references to those specific

sources and documents of which the petitioner is aware and on which the

petitioner intends to rely to establish those facts or expert opinion.

Petitioner must provide sufficient information to show that a genuine

dispute exists with the applicant on a material issue of law or fact.

Contentions shall be limited to matters within the scope of the

amendment under consideration. The contention must be one which, if

proven, would entitle the petitioner to relief. A petitioner who fails

to file such a supplement which satisfies these requirements with

respect to at least one contention will not be permitted to participate

as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

operator should be given Datagram Identification Number N1023 and the

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

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

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

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

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

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

North Carolina

Date of amendment request: September 18, 1996

Description of amendment request: Revise Technical Specification

(TS) 4.8.1.1.2 by removing TS 4.8.1.1.2.h.2 pressure testing

requirement since adequate testing will be completed in accordance with

American Society of Mechanical Engineers (ASME) Boiler and Pressure

Vessel Code, Section XI.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

This change does not involve a significant hazards consideration

for the following reasons:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Applying ASME Code, Section XI alternative examination/testing

will not affect any initiators of any previously evaluated accidents

or change the manner in which the emergency diesel generators or any

other systems operate. The diesel fuel oil system supports the

emergency diesel generators which serve an accident mitigating

function. Where portions of piping are non-isolable or where

atmospheric tanks are involved, the Section XI ASME alternatives to

110% pressure testing continue to ensure the integrity of the fuel

oil system without any impact on analyzed accident scenarios or

their consequences. Therefore, the proposed amendment does not

result in an increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed alternative testing and surveillance will not

involve any physical alterations or additions to plant equipment or

alter the manner in which any safety-related system performs it

function. Using ASME Section XI, or NRC-approved ASME Code cases, as

guidance for pressure testing continues to provide assurance that

the fuel oil supply system will perform its intended function.

Therefore, the proposed changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

There are no changes being made to the safety limits or safety

settings that would adversely impact plant safety. Further, there is

no impact on the margin of safety as defined in the Technical

Specifications. Utilizing ASME Section XI as guidance for

determining those sections of piping that should be pressure-tested

or tested at atmospheric pressure will ensure proper operation of

the diesel generator fuel oil supply system. Therefore, the proposed

changes do not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

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

Raleigh, North Carolina 27602

NRC Project Director: F. Mark Reinhart, Acting

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: August 29, 1996 (Reference NRC-96-0111)

Description of amendment request: The proposed amendment will: (1)

allow certain equipment and instruments to be removed from service for

short periods of time to allow for

[[Page 55030]]

maintenance, testing, inspection, modifications, and account for

equipment failures; (2) reduce the frequency of environmental liquid

effluent monitoring and eliminate one raw water sampling location; (3)

eliminate the requirement for moisture intrusion monitoring for the

reactor building lower level; and (4) correction of a typographical

error.

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 using the standards in 10 CFR 50.92(c). The licensee's

analysis is presented below:

(1) The operation of Enrico Fermi Atomic Power Plant, Unit 1, in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes do not involve a significant increase in

the probability or consequences of an accident. Provisions for

removing the primary cover gas supply from service for short periods

of time will not significantly increase the probability of an

accident occurring as long as the probability of a significant water

reaction with residual sodium is not significantly increased. This

is ensured by prescribing limits on the time that carbon dioxide

pressure can be low. The consequences of an accident would not be

affected by provisions for removing the primary cover gas supply

from service as this equipment does not mitigate accidents or affect

the accident sequences. Similarly, the provisions for removing the

moisture intrusion and cover gas pressure alarms from service for

short period of time will not significantly increase the probability

of an accident. The alarms provide a monitoring function to detect

degradation in the performance of the cover gas supply and sump

systems. Absence of these alarm functions for short periods of time

does not increase the probability of such degradation and it does

not significantly impact the ability for timely detection of such

degradation. The consequences of an accident would not be affected

by provisions for removing the moisture intrusion and cover gas

pressure alarms from service as this equipment does not mitigate

accidents or affect the accident sequences. Elimination of the

moisture intrusion alarm for the reactor building lower level does

not significantly increase the probability of an accident because

the probability that water could accumulate in this area is

essentially unchanged. Design features of the foundation,

containment structure, and annulus drains are intended to prevent

entry of water into the reactor building. These features have

prevented any water intrusion into this area. The consequences of an

accident would not be affected by elimination of the moisture

intrusion alarm for the reactor building lower level because this

equipment does not mitigate accidents or affect the accident

sequences. The Safety Evaluation Supporting Amendment 9 to the

referenced license did not rely on moisture intrusion monitoring and

alarm features for any safety function or accident prevention or

mitigation function. Environmental monitoring surveillance are

unrelated to postulated accident sequences and cannot affect the

probability or consequences of an accident. The correction of the

typographical error is unrelated to accident initiation and

sequences and cannot affect the probability or consequences of any

accident.

(2) The operation of Enrico Fermi Atomic Power Plant, 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.

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

different accident from any previously evaluated. With the exception

of the allowance for composite environmental samples, which are

unrelated to any potential accident sequence, these changes propose

no new activities or new methods for performing existing activities.

Previous evaluations have considered the release of all of the

radioactivity in the residual sodium due to postulated fire or other

catastrophe and release of radioactive water stored in the liquid

waste tanks which bound the only possible radiological accidents at

Fermi 1. For these reasons, no new or different type of accident is

created by these changes.

(3) The operation of Enrico Fermi Atomic Power Plant, Unit 1, in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

The proposed changes do not involve a significant reduction in a

margin of safety. The changes to the primary system cover gas system

technical specifications still ensure that any residual sodium is

passivated by carbon dioxide. Changes to the alarms affect only

monitoring functions and therefore do not cause a change to any

parameter that could affect the margin of safety. Similarly, the

environmental surveillances are unrelated to margin of safety. The

correction of the typographical error is unrelated to margin of

safety. For these reasons, the proposed changes do not involve a

significant reduction in the margin of safety.

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

this review, it appears that the three standards of 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, Esquire, Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226NRC Branch Chief: Michael F.

Weber

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

Michigan

Date of amendment request: September 25, 1996 (NRC-96-0085)

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Surveillance Requirement 4.8.4.3 to

remove the requirement to periodically test the thermal overload (TOL)

devices for safety-related motor-operated valves (MOVs). The

surveillance requirement would continue to require testing of a TOL

device following any maintenance activity that could affect the

performance of the device. The surveillance requirement would also be

clarified by indicating that testing of TOL devices is required upon

initial installation. The associated portion of the TS Bases would also

be revised to reflect this change.

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. The deletion of

the requirement for testing of the TOL protective devices lessens

degradation to the components which can improve MOV reliability.

Based on historical data through the years of testing, there is no

significant drifting of the trip setpoints of the TOL protective

devices. The probability of an accident would not increase since

terminating the periodic testing or clarifying the situational

testing requirements cannot cause equipment to operate inadvertently

and so cannot cause an accident. The periodic testing of the TOL

protective devices can temporarily render MOVs inoperable due to the

removal of the components from service and can cause safety systems/

divisions to become unavailable. The deletion of the periodic

testing requirement would increase the availability of safety

systems insuring that they would be able to respond to accident

conditions. The consequences of an accident will not increase since

eliminating the periodic testing and clarifying the situational

testing requirements will improve reliability of safety-related MOVs

to respond to an accident and will not increase the failure rate of

equipment. The clarification of the situational testing ensures that

the test will be conducted after any maintenance that could affect

the performance of the TOL protective devices. Thus, the proposed

change increases reliability of the MOVs and increases plant safety.

Therefore this change will not result in a significant increase in

the probability or consequences of an accident.

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

or different accident from any previously evaluated. The TOL

[[Page 55031]]

protective devices are not an accident initiator, they only protect

equipment provided to mitigate the consequences of an accident. For

this reason, no new or different type of accident is created by this

change.

3. The proposed change does not involve a significant reduction

in a margin of safety. The trip setpoints of the TOL protective

devices depend upon both the current and the length of time the

current is applied. The trip setpoints for TOL protective devices

are much higher than conditions normally experienced during an MOV

stroke and are meant to protect the motor from stall and overload

conditions. The difference between the current of the trip setpoints

and the normal conditions is great enough that a premature trip of

the TOL protective device is highly unlikely, even at degraded

voltages. The TOL protective device protects the motor from the

stall conditions. Not conducting the periodic testing of the TOL

protective devices would not cause the MOVs to fail, nor would the

performance of the MOVs be adversely affected. Throughout the life

of the plant, there has never been an instance of a safety related

MOV failure due to degradation or failure of TOL protective devices.

Further, based on maintenance history, the elimination of the

periodic testing would eliminate any significant potential

degradation of the TOL protective devices, thereby increasing their

reliability. Finally, with the removal of the periodic testing of

the TOL protective devices, fewer MOVs would have to be removed from

service for testing. Since necessary components would no longer be

inoperable due to the periodic testing, there would be an increase

of availability time of safety systems/divisions. Deletion of the

periodic testing could reduce the durations of online system

outages. Clarifying the situational testing requirements would

better define when the testing of the TOL protective devices is

necessary which would ensure operability. The testing would be based

on installation or any maintenance that could affect the TOL

protective device. For these reasons, the proposed change does not

involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

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

2000 Second Avenue, Detroit, Michigan 48226

NRC Project Director: John N. Hannon

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

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

Date of amendment request: June 21, 1996

Description of amendment request: The proposed amendments would

administratively correct the term ``lifting load'' in Technical

Specification 3.9.6b.2 to ``lifting force.'' This correction would

clarify that the static loads associated with the lifting tool, drive

rod and control rod weights are not included in the lifting force

limit. The amendments would also more accurately define auxiliary hoist

minimum capacities and give a more expansive description of the

activities for which protective measures and surveillance testing are

used.

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:

Question: Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The proposed change[s] [are] administrative in nature, and

do[] not represent any changes to the refueling process in the

field. It more accurately describes the components for which the

LCO's [limiting conditions of operation] protection is intended as

well as giving a more accurate description of the auxiliary hoist's

minimum capacity. [They] also broaden[] the domain of activities for

which protective measures are taken, by including drag load testing

into monitored activities. At both MNS [McGuire Nuclear Station] and

CNS [Catawba Nuclear Station], the auxiliary hoists and the

manipulator cranes are rated at [greater than or equal to] 3000

pounds and are surveillance tested to greater than 1000 pounds. This

brackets the limit force lifting value change from 600 to 1000

pounds in the amendment proposal.

Question: Will the change create the possibility of a new or

different kind of accident from any accident previously evaluated?

No. Th[ese] proposed administrative change[s] reflect[] no

changes in the refueling processes, or any systems, structures or

components connected with the refueling process.

Question: Will the change involve a significant reduction in a

margin of safety?

No. The proposed administrative change[s] [have] no impact on

refueling processes, systems, structures or components, and do[] not

result in any significant reduction in a margin of safety. The

subject change[s] only clarif[y] the original intent of the

specification and more accurately describe[] the involved

components, component capacities and the domain of activities for

which measures are taken to protect the reactor internals.

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

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

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

proposed amendments involve no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina

Date of amendment request: September 17, 1996 (TSC 96-01)

Description of amendment request: The proposed changes would reduce

the Reactor Building pressure setpoint for actuation of the Reactor

Building Spray System in Technical Specification (TS) 3.5.3 from a

maximum of 30 pounds per square inch gauge (psig) to 15 psig, reduce

the maximum allowable Reactor Building internal pressure specified in

TS 3.6.4 from 1.5 psig to 1.2 psig when the reactor is critical, revise

the corresponding Bases of TS 3.3 to indicate that the Reactor Building

sprays and coolers are designed to mitigate the containment temperature

response rather than containment pressure response to a loss-of-coolant

accident, and make other administrative changes. In addition, the lower

Reactor Building pressure limit (a vacuum of 5 inches of mercury (Hg))

in Specification 3.6.4 would be changed to the corresponding value in

terms of psig to reflect the units displayed on the control room

instrumentation.

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:

No. The analysis of the post-LOCA [loss-of-coolant accident]

Reactor Building response to high-energy line breaks, using the new

methodology, uses assumptions different from the requirements

currently delineated in Technical Specifications. The new

assumptions used for initial Reactor Building pressure and Reactor

Building Spray system

[[Page 55032]]

actuation are 1.2 psig and 20 psig respectively. These values are

lower, and hence more conservative, than the values currently

specified in Technical Specifications.

Since the new values for Reactor Building pressure and Reactor

Building Spray actuation are more conservative and the analysis

methodology has received approval from the NRC via [an] SER, this

change does not involve a significant increase in the probability or

consequences of an accident previously identified.

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

accident from any kind of accident previously evaluated:

No. The methodology for Reactor Building high energy line break

analysis is being revised. The revision of the method of analysis

does not alter the manner by which plant systems and components

function for accident mitigation.

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

No. By letter dated March 15, 1995, the NRC stated that the new

analyses described in the topical report, DPC-NE-3003-P, expand the

scope of analyzed piping failures in containment for the Oconee

facilities. The NRC further stated that this new analysis method has

been used to reanalyze existing licensing basis pipe failure events

in containment, and to examine the potential effects of previously

unanalyzed assumptions and initial conditions which the NRC staff

finds to be consistent with current NRC staff acceptance criteria or

produce equally conservative results. In conclusion, the NRC

confirmed that this methodology, with appropriate adjustments to

reflect potential plant modifications, may be used by Duke Power to

perform future analyses in support of licensing applications related

to containment accident response. This proposed change to Technical

Specifications reflects the use of this new methodology. Based on

this new methodology, changes have been made to setpoint assumptions

for initial Reactor Building pressure and Reactor Building Spray

actuation. This proposed Technical Specification change reflects

those assumption changes. This methodology has been accepted by the

NRC. This proposed change to Technical Specifications does not

involve a significant reduction in the margin of safety.

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

thisreview, 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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley

Power Station, Unit No. 1, Shippingport, Pennsylvania

Date of amendment request: September 9, 1996

Description of amendment request: The proposed amendment would

revise the Minimum Channels Operable requirement of Item 4.c (Steam

Line Isolation, Containment Pressure Intermediate -- High-High) of

Technical Specification (TS) Table 3.3-3 from 3 to 2. This proposed

change would make this Unit 1 TS consistent with the comparable Unit 2

TS.

The proposed amendment would also revise the minimum charging pump

discharge pressure in TS 3.5.5 from 2311 psig to 2397 psig. This change

is required to ensure that safety analysis assumptions for safety

injection flow are met. Conforming changes would also be made to the

Bases for TS 3/4.5.5 to reflect the proposed changes to TS 3.5.5.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed amendment does not add or modify any existing plant

equipment. Since normal charging pump discharge pressure is greater

than or equal to approximately 2440 psig, no additional plant

configuration changes or modifications will be required to comply

with this revised charging pump discharge pressure value. The

proposed amendment does not change the design or function of the

containment pressure intermediate-high-high channels.

The consequences of an accident previously evaluated are not

significantly increased. The ability of the containment pressure

intermediate-high-high function to initiate steam line isolation

will not be affected. Since steam line isolation will continue to

occur at the same required trip setpoint, the amount of mass and

energy released to containment along with the ability to maintain at

least one unfaulted steam generator (SG) as a heat sink for the

reactor remains unchanged. The amount of seal injection flow will

continue to be adequately limited to ensure sufficient flow to the

reactor core during accident conditions. The Bases changes are

editorial in nature and do not involve a change to probability or

consequences of an accident previously evaluated.

Based on the above discussion, it is concluded that this change

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed amendment does not change the plant configuration

in a way which introduces a new potential hazard to the plant. Since

design requirements continue to be met and the integrity of the

reactor coolant system pressure boundary is not challenged, no new

failure mode has been created. As a result, an accident which is

different than already evaluated in the Updated Final Safety

Analysis Report will not be created due to this change.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

of safety?

The margin of safety is not significantly reduced by this

proposed change. The trip setpoint for the containment pressure

intermediate-high-high function remains unchanged. With one channel

inoperable, the remaining two channels will continue to initiate the

protective function on a two-out-of-two logic. The action statement

limits this condition to 6 hours after which time the inoperable

channel must be placed in the trip condition. This action restores

the function to be able to meet single failure criteria on a one-

out-of-two logic basis.

The proposed revision to the charging pump discharge pressure

will not change the flow limit on seal injection. The specification

will continue to ensure that seal injection flow is limited. This

will ensure that sufficient flow to the reactor core is provided

during accident conditions.

The proposed changes to the Bases for seal injection flow are

editorial in nature and do not affect the margin of safety.

Therefore, this proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001

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

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

NRC Project Director: John F. Stolz

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: August 29, 1996

Description of amendment request: The proposed amendment would

revise the technical specifications (TSs) to reflect the elimination of

T-factor adjustments in the Average Power

[[Page 55033]]

Range Monitors (APRM) setpoints, a decrease in the calibration

frequency of the Local Power Range Monitors (LPMR), and an improvement

in the calculation of Reactivity Anomaly.

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 request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

This change replaces the APRM setpoints T-factor limit with

power and flow-dependent minimum critical power ratio (MCPR) and

linear heat generation rate (LHGR) limits. These new power and flow-

dependent thermal limits eliminate the need for manual setpoint

adjustment resulting from power peaking conditions. The new power

and flow-dependent thermal limits are automatically applied by

computer software during the calculation of the core thermal limits

and, therefore, do not require manual setpoint adjustments based on

the power peaking conditions in the reactor. Extensive transient

analyses at a variety of power and flow conditions have been

performed and were utilized to study the trend of transient severity

without the setpoints T-factor limit. A large data base was

established by analyzing limiting transients over a range of power

and flow conditions. The data base included evaluations

representative of a variety of plant configurations and parameters

such that the conclusions drawn from the studies would be applicable

to the broad range of boiling water reactors (BWRs). This data base

was utilized to develop plant specific operating limits (MCPR and

LHGR), which assures that margins to fuel safety limits are equal to

or larger than those currently in existence with the APRM setpoints

T-factor limit applied. Therefore, this change does not involve an

increase in the probability of any event previously evaluated.

The consequences of an accident previously evaluated have not

been increased because, in all cases, the new power and flow-

dependent thermal limits (MCPR and LHGR) assure that margins to fuel

safety limits are equal to or larger than those currently in

existence with the APRM setpoints T-factor limit applied. Protection

of other thermal limits for all previously analyzed events is

accomplished by specific limits that are independent of the APRM

setpoints T-factor. These are the power and flow-dependent MCPR

Operating Limits which provide protection from fuel dryout and the

rated maximum average planner linear heat generation rate (MAPLHGR)

limit which provides protection of the peak clad temperature for the

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

Therefore, the proposed change does not involve a significant

increase in the consequences of any event previously evaluated.

No new equipment is introduced by the change in the local power

range monitor (LPRM) calibration frequency and, therefore, the

probability for an accident previously evaluated is unchanged. The

consequences of an accident can be affected by the thermal limits

prior to the accident but LPRM chamber and cycle exposure have no

significant effect on the calculated thermal limits. The thermal

limit calculation is not significantly effected because the LPRM

sensitivity versus exposure function is well defined. This allows

accurate LPRM end-of-life calculations so that detectors can be

replaced before their behavior significantly deteriorates. In the

event deterioration is noted late in the cycle for a few chambers,

they can be bypassed with no significant effect on uncertainties.

Also, the total nodal power uncertainty remains less than the

uncertainty assumed in the General Electric BWR Thermal Analysis

Basis (GETAB) safety limit. Therefore, the thermal limit calculation

is not affected by the LPRM calibration frequency and the

consequences of an accident previously evaluated are not changed.

The change in the parameters used to measure reactivity for

calculation of the reactivity anomaly has no affect on either the

consequences or the probability of an accident previously evaluated

because the allowed reactivity anomaly criteria is unchanged. The

only change is the parameters used to measure reactivity.

Therefore, the proposed elimination of the APRM setpoints T-

factor maintains adequate off-rated MCPR and LHGR margin for all

operating conditions. Also, the change in the LPRM calibration

frequency continues to maintain the accuracy of the thermal limit

calculation. Therefore, the consequences of an accident previously

evaluated are not affected by this change. Finally, the change in

the parameters used to measure reactivity for calculation of the

reactivity anomaly has no affect on either the consequences nor the

probability of an accident previously evaluated. Since no new plant

equipment is introduced by any of the proposed changes, the

probability of accidents previously evaluated are not changed.

Therefore, none of the proposed changes involve an increase in the

probability or consequences of any event previously evaluated.

2. The request does not create the possibility of occurrence of

a new or different kind of accident from any accident previously

evaluated.

This change only replaces the APRM setpoints T-factor limit with

power and flow-dependent MCPR and LHGR limits, changes the LPRM

calibration frequency, and a change to the parameter(s) used to

measure reactivity. None of the proposed changes involve any new

modes of operation or any plant modifications. Therefore, the

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

type of accident from any accident previously analyzed.

3. The request does not involve a significant reduction in a

margin of safety.

The replacement of the APRM setpoints T-factor limit with power

and flow-dependent thermal limits has been confirmed to provide

adequate MCPR and LHGR protection at all reactor operation

conditions. Operation with higher peaking without APRM gains or flow

bias trip setpoints adjustment does not involve a reduction in a

margin of safety because the higher power peaking resulting from

elimination of the APRM setpoints T-factor has been analyzed to

assure that the margins to fuel safety limits are equal to or larger

than those currently in existence with the APRM setpoints T-factor

limit applied. Therefore, the replacement of the APRM setpoint T-

factor with power and flow-dependent thermal limits does not involve

a reduction in the margin of safety.

Protection of other thermal limits for all previously analyzed

events is accomplished by specific limits that are independent of

the APRM setpoint T-factor limit. These are the power and flow-

dependent

MCPR Operating Limits which provide protection from fuel dryout

and the rated MAPLHGR limit which provides protection of the peak

clad temperature for the DBA LOCA.

The margin of safety can be affected by the thermal limits prior

to an accident but LPRM chamber exposure and cycle exposure have no

significant effect on the calculated thermal limits. The thermal

limit calculation is not significantly affected because the LPRM

sensitivity versus exposure function is well defined. This allows

accurate LPRM end of life calculations so that detectors can be

replaced before their behavior significantly deteriorates. In the

event deterioration is noted late in the cycle for a few chambers,

they can be bypassed with no significant effect on uncertainties.

Also, the total nodal power uncertainty remains less than the

uncertainty assumed in the GETAB safety limit. Therefore neither the

thermal limit calculation nor the margin of safety are affected by

the LPRM calibration.

The change in the parameters used to measure reactivity for

calculation of the reactivity anomaly has no affect on the margin of

safety because the allowed reactivity anomaly criteria is unchanged.

The only change is the parameters used to measure reactivity.

Neither the change to APRM setpoints T-factor nor the change to

the LPRM calibration frequency significantly effects the thermal

limits calculation, and, therefore, do not result in an increase in

core damage frequency. The change in the parameters used to measure

reactivity for calculation of the reactivity anomaly has no affect

on the core damage frequency because the allowable reactivity

anomaly criteria remains unchanged. Therefore, the proposed changes

do not involve a reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Documenmt Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

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

[[Page 55034]]

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

NRC Project Director: William D. Beckner

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: August 29, 1996

Description of amendment request: The proposed amendment would

provide a revision to the reactor pressure vessel (RPV) surveillance

capsule withdrawal schedule for the River Bend Station. The first

surveillance capsule would be withdrawn at 10.4 effective full power

years (EFPY) rather than at 6EFPY.

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.

Pressure-temperature (P-T) limits (RBS Technical Specifications

Figure 3.4.11-1) are imposed on the reactor coolant system to ensure

that adequate safety margins against nonductile or rapidly

propagating failure exist during normal operation, anticipated

operational occurrences, and system hydrostatic tests. The P-T

limits are related to the nil-ductility reference temperature,

RTNDT, as described in ASME Section III, Appendix G. Changes in

the fracture toughness properties of RPV beltline materials,

resulting from the neutron irradiation and the thermal environment,

are monitored by a surveillance program in compliance with the

requirements of 10CFR50, Appendix H. The effect of neutron fluence

on the shift in the nil-ductility reference temperature of pressure

vessel steel is predicted by methods give in Regulatory Guide 1.99,

Rev. 2.

River Bend's current P-T limits were established based on

adjusted reference temperatures developed in accordance with the

procedures prescribed in Reg. Guide 1.99, Rev. 2, Regulatory

Position 1. Calculation of adjusted reference temperature by these

procedures includes a margin term to ensure conservative, upper-

bound values are used for the calculation of the P-T limits.

Revision of the first capsule withdrawal schedule will not affect

the P-T limits because they will continue to be established in

accordance with Regulatory Position 1 (or other NRC-approved)

procedures. When permitted (two or more credible surveillance data

sets available), Regulatory Position 2 (or other NRC-approved)

methods for determining adjusted reference temperature will be

followed.

This change is not related to any accidents previously

evaluated. The proposed change is a revision of the Withdrawal Time

for the first surveillance capsule as given in Technical

Requirements (TR) Table 3.4.11-1 from 6 EFPY to 10.4 EFPY. This

change will not affect P-T limits as given in RBS Technical

Specifications Figure 3.4.11-1 or USAR Figures 5.3-4a and 5.3-4b.

This change will not affect any plant safety limits or limiting

conditions of operation. The proposed change will not affect reactor

pressure vessel performance as no physical changes are involved and

RBS vessel P-T limits will remain conservative in accordance with

Reg. Guide 1.99, Rev. 2 requirements. The proposed change will not

cause the reactor pressure vessel or interfacing systems to be

operated outside of their design or testing limits. Also, the

proposed change will not alter any assumptions previously made in

evaluating the radiological consequences of accidents. Therefore,

the probability or consequences of accidents previously evaluated

will not be increased by the proposed change.

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

or different kind of accident from any accident previously

evaluated.

The proposed change is a revision of the Withdrawal Time in TR

Table 3.4.11 for the first RPV material surveillance capsule from 6

EFPY to 10.4 EFPY. This proposed change does not involve a

modification of the design of plant structures, systems, or

components. The proposed change will not impact the manner in which

the plant is operated as plant operating and testing procedures will

not be affected by the change. The proposed change will not degrade

the reliability of structures, systems or components important to

safety (ITS) as equipment protection features will not be deleted or

modified, equipment redundancy or independence will not be reduced,

supporting system performance will not be downgraded, the frequency

of operation of ITS equipment will not be increased, and increased

or more severe testing of ITS equipment will not be imposed. No new

accident types or failure modes will be introduced as a result of

the proposed change. Therefore, the proposed change does not create

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

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

As stated in the River Bend SER, ``Appendices G and H of 10CFR50

describe the conditions that require pressure-temperature limits and

provide the general bases for these limits. These appendices

specifically require that pressure-temperature limits must provide

safety margins at least as great as those recommended in the ASME

Code, Section III, Appendix G. .... Until the results from the

reactor vessel surveillance program become available, the staff will

use RG 1.99, Revision 1 [now Revision 2] to predict the amount of

neutron irradiation damage. ... The use of operating limits based on

these criteria--as defined by applicable regulations, codes, and

standards--will provide reasonable assurance that nonductile or

rapidly propagating failure will not occur, and will constitute an

acceptable basis for satisfying the applicable requirements of GDC

31.''

Bases for RBS Technical Specification 3/4/11 states: ``The P/T

limits are not derived from Design Basis Accident (DBA) analyses.

They are prescribed during normal operation to avoid encountering

pressure, temperature, and temperature rate of change conditions

that might cause undetected flaws to propagate and cause nonductile

failure of the RCPB [Reactor Coolant Pressure Boundary], a condition

that is unanalyzed. ... Since the P/T limits are not derived from

any DBA, there are no acceptance limits related to the P/T limits.

Rather, the P/T limits are acceptance limits themselves since they

preclude operation in an unanalyzed condition.''

The proposed change will not affect any safety limits, limiting

safety system settings, or limiting conditions of operation. The

proposed change does not represent a change in initial conditions,

or in a system response time, or in any other parameter affecting

the course of an accident analysis supporting the Bases of any

Technical Specification. The proposed change does not involve

revision of the P-T limits but rather a revision of the Withdrawal

Time for the first surveillance capsule. The current P-T limits were

established based on adjusted reference temperatures for vessel

beltline materials calculated in accordance with Regulatory Position

1 of Reg. Guide 1.99, Rev. 2. P-T limits will continue to be revised

as necessary for changes in adjusted reference temperature due to

changes in fluence according to Regulatory Position 1 until two or

more credible surveillance data sets become available. When two or

more credible surveillance data sets become available, P-T limits

will be revised as prescribed by Regulatory Position 2 of Reg. Guide

1.99, Rev. 2 or other NRC-approved guidance. Therefore, the proposed

changes do not involve a significant reduction in any margins of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

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

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

NRC Project Director: William D. Beckner

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

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

Date of amendment request: September 23, 1996

Description of amendment request: The proposed amendment would

revise the Crystal River Unit 3 (CR 3) technical specifications (TS) to

delete a note

[[Page 55035]]

associated with Surveillance Requirement (SR) 3.3.7.1 for the

Engineered Safeguard Actuation System (ESAS) Automatic Actuation Logic.

Applicable TS Bases will also be revised to reflect the proposed TS

change.

SR 3.3.7.1 requires periodic testing of the ESAS automatic

actuation logic matrix to demonstrate that the required logic

combinations are operable. When the ESAS automatic actuation logic is

placed in an inoperable status solely for performing of this

surveillance, the note associated with the SR 3.3.7.1 provides relief

in that it allows not entering into applicable Conditions and Required

Actions for up to 8 hours, provided the associated engineering

safeguards (ES) function is maintained. The licensee has determined

that because of the CR 3 design of the ESAS System and the way the test

is performed, maintenance of the ``associated ES function'' is not

possible. Thus, the note does not provide the relief intended and

therefore, the licensee proposes to delete the note. During the

performance of the ESAS test and bypassing the associated ES function,

the licensee proposes to enter into applicable TS Conditions.

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 will not significantly increase the

probability or consequences of an accident previously evaluated

because unavailability of equipment is recognized in the design of

the plant and in the Technical Specifications. The probability and

consequences of accidents previously evaluated are bounded by the

evaluations done for the allowed outage time of the associated

functions.

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

or different kind of accident from any accident previously evaluated

because the bypassing of ES functions for testing purposes does not

place the plant in a configuration which would allow the possibility

of a new or different kind or accident to be created.

3. The proposed change will not involve a significant reduction

to the margin of safety because deleting the NOTE does not effect

the way the test is performed. The test is required by the Technical

Specifications and will still be performed in the same manner. Thus,

there is no change in the unavailability of the system as a result

of this change and the margin of safety is not reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

Attorney for licensee: A. H. Stephens, General Counsel, Florida

Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida

33733

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: September 27, 1996

Description of amendment request: The proposed amendment would

revise the Crystal River 3 (CR3) post-accident monitoring (PAM)

instrumentation technical specification (TS). Specifically, the

following TS changes are proposed:

A. Table 3.3.17-1, Function 8: The descriptor is changed from

``Containment Pressure (Narrow Range)'' to ``Containment Pressure

(Expected Post-Accident Range).''

B. Table 3.3.17-1, Function 18: The required channels for Core Exit

Temperature (Backup) is changed from ``2 sets of 5'' to ``3 per core

quadrant.''

C. Table 3.3.17-1: A new Function 20 is added and designated as

``Low Pressure Injection Flow.''

D. Table 3.3.17-1: A new Function 21 is added and designated as

``Degrees of Subcooling.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration (the letters A, B, C and D correspond to the proposed TS

changes), which is presented below:

1. The proposed changes will not significantly increase the

probability or consequences of an accident previously evaluated

because:

A/B. The changes in containment pressure and core exit

thermocouple nomenclature do not reflect any physical changes to the

facility.

C/D.The addition of low pressure injection flow and degrees of

subcooling to the Post-Accident Monitoring Instrumentation LCO is

being done to comply with a commitment made during the technical

specification improvement program to include in the technical

specifications, that instrumentation which monitors variables

classified as Type A in accordance with Regulatory Guide 1.97. These

two variables have recently been re-classified as Type A. The

associated instruments are used after an accident occurs to prompt

the operators to take certain mitigative actions. Therefore, the

probability of an accident occurring is unaffected. As part of the

re-classification of these variables to Type A, the associated

monitoring instrumentation will be under more strict surveillance

and control, which provides additional assurance that the prescribed

manual operator actions will be implemented when necessary. This, in

turn, assures the previously evaluated accident consequences remain

valid.

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

or different kind of accident from any accident previously evaluated

because:

A/B. The changes in containment pressure and core exit

thermocouple nomenclature do not reflect any physical changes to the

facility. The changes provide clarification for the instruments

which are required to comply with the LCO.

C/D. The addition of low pressure injection flow and degrees of

subcooling to the Post-Accident Monitoring instrumentation LCO is

being done to comply with a commitment made during the technical

specification improvement program to include in the technical

specifications, that instrumentation which monitors variables

classified as Type A in accordance with Regulatory Guide 1.97. These

two variables have been re-classified as Type A. The associated

instruments are used after an accident occurs to prompt the

operators to take certain mitigative actions. Since the

instrumentation is used only post-accident, these changes do not

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

3. The proposed change will not involve a significant reduction

to the margin of safety because:

A/B. The changes in containment pressure and core exit

thermocouple nomenclature have no affect on the margin of safety.

The changes provide clarification of the technical specifications.

This reduces the potential for confusion regarding this

instrumentation.

C/D. The addition of low pressure injection flow and degrees of

subcooling to the post-accident monitoring instrumentation table

adds controls on the OPERABILITY of post-accident monitoring

instrumentation providing greater assurance it will be available

should an accident occur.

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: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

Attorney for licensee: A. H. Stephens, General Counsel, Florida

Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida

33733

NRC Project Director: Frederick J. Hebdon

[[Page 55036]]

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

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

Connecticut

Date of amendment request: September 5, 1996

Description of amendment request: The proposed change deletes

License Condition 2.C.5, Integrated Implementation Schedule.

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:

In accordance with 10CFR50.92, NNECO has reviewed the attached

proposed change and has concluded that it does not involve a

significant hazards consideration (SHC). The basis for this is that

the three criteria of 10CFR50.92(c) are not compromised. The

proposed change does not involve an SHC because the change would

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Operation of the facility in accordance with the proposed change

would result in a change in an administrative process for

prioritizing and scheduling projects and engineering evaluations.

With the limited number of NRC required projects remaining to be

implemented, the IIS [Integrated Implementation Schedule] is no

longer required to schedule resources for the remaining topics.

Since this license condition only involves an administrative

process, it does not directly affect the design or operation of the

plant. Therefore, no accident analyses are affected by the change,

and the change does not increase the probability or consequences of

any previously evaluated accident.

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

from any previously evaluated.

The proposed license modification removes a requirement relating

to the scheduling of modifications and engineering evaluations.

Because the license condition addresses only an administrative

scheduling mechanism, it does not affect directly the design or

operation of the plant. Therefore, the proposed change does not

create a different kind of accident from those previously analyzed.

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

The proposed license modification removes a requirement relating

to the scheduling of modifications and engineering evaluations. The

original purpose of the IIS and the ISAP [Integrated Safety

Assessment Program] was to prioritize and schedule modifications and

engineering evaluations in a manner that was agreed upon by both

NNECO and the NRC. These programs were especially important to

Millstone Unit No. 1 for priorization of topics associated with the

SEP [Systematic Evaluation Program] and the TMI [Three Mile Island]

Action Plan. This program is considered to be no longer necessary.

Modifications and engineering evaluations will be scheduled and

prioritized using other methodologies. Since this change involves an

administrative process only, there is no direct impact on the design

or operation of the plant, and therefore, no 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: Learning Resources Center,

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

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

Rope Ferry Road, Waterford, CT 06385

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

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

CT 06141-0270

NRC Project Director: Phillip F. McKee

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

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

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

Date of application for amendments: August 27, 1996

Description of amendment request: The proposed amendment revises

the required value of control rod drive (CRD) system pressure in

technical specification (TS) 3.10.8, ``Shutdown Margin (SDM) Test-

Refueling.''

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 any accident previously

evaluated.

The proposed changes are purely administrative and do not

involve any physical changes to plant SSC [systems, structures and

components]. The change in the minimum CRD charging water header

pressure from 955 psig to 940 psig was previously approved in TS

Amendments Nos. 211 and 216 for PBAPS [Peach Bottom Atomic Power

Station], Units 2 and 3. TS Change Request 95-12 was incomplete by

inadvertently failing to identify the need to change requirement (f)

of LCO [Limiting Condition for Operation] 3.10.8. Therefore, the

proposed changes will not increase the probability of occurrence or

the consequences of an accident previously evaluated in the SAR

[safety analysis report].

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 purely administrative and do not

involve any physical changes to plant SSC. The proposed changes do

not allow plant operation in any mode that is not already evaluated

in the SAR. Therefore, the possibility of a different type of

accident than previously evaluated in the SAR is not created.

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

in the margin of safety.

The proposed changes are purely administrative and have no

impact on any safety analysis assumptions or margins of safety. A

change to SR 3.10.8.6 was approved by the NRC by TS Amendment Nos.

211 and 216. LCO 3.10.8 requirement (f) should have been changed at

the same time to reflect a minimum CRD charging water pressure of

940 psig. Changing LCO 3.10.8 requirement (f) to reflect TS

Amendment Nos. 211 and 216 is purely administrative, and therefore,

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: Government Publications

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

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

PA 17105

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

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

PA 19101

NRC Project Director: John F. Stolz

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

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

Pennsylvania

Date of amendment request: May 20, 1996

Description of amendment request: The proposed Technical

Specifications (TS) changes would revise TS Sections 3/4.4.9.2, 3/

4.9.11.1, 3/4.9.11.2, and the associated TS Bases 3/4.4.9 and 3/4.9.11,

to more clearly describe that the Residual Heat Removal (RHR) system

Shutdown Cooling mode of operation consists of four (4) ``subsystems.''

These TS sections pertain to plant operations during Operational

Conditions (OPCONs) 4, ``Cold Shutdown'' and 5, ``Refueling.'' In

addition, the proposed TS change would make administrative changes to

TS Section 3/4.4.9.1 to

[[Page 55037]]

ensure consistency in terminology regarding the description of Shutdown

Cooling ``subsystems.'' The proposed TS changes are consistent with the

guidance delineated in the Improved TS (i.e., NUREG-1433, Revision 1,

``Standard Technical Specifications General Electric Plants, BWR/4,''

dated April 1995) which indicates that the RHR Shutdown Cooling mode of

operation is comprised of two (2) loops and four (4) subsystems (i.e.,

two (2) subsystems per loop).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS changes do not involve any physical changes to

plant structures systems, or components. The RHR [Residual Heat

Removal] Shutdown Cooling mode of operation is manually controlled

and is not required for accident mitigation. The RHR system will

continue to function as designed in all modes of operation. The

consequences of equipment malfunction are not changed from those in

existing analyses, with no increase in onsite or offsite

radiological effects. The RHR system will continue to function as

designed to mitigate the consequences of an accident and resultant

onsite and offsite radiological effects remain as previously

evaluated. The proposed TS changes will revise the TS to more

clearly describe the RHR system configuration in OPCONs 4 and 5. The

proposed changes are consistent with the guidance stipulated in

NUREG-1433, Revision 1.

The four (4) ``subsystem'' Shutdown Cooling designation permits

operability of only one (1) RHR heat exchanger for Shutdown Cooling

service in Operational Conditions (OPCONs) 4 and 5, as long as both

associated RHR pumps are operable and alignable for Shutdown

Cooling. TS requirements for RHR Shutdown Cooling operation in Hot

Shutdown, Suppression Pool Spray, and Suppression Pool Cooling

continue to require two (2) independent loops to be operable in

OPCONs 1, 2, and 3*, meaning both RHR heat exchangers will still be

required to be operable throughout OPCON 3.

The four (4) ``subsystem'' Shutdown Cooling designation has no

effect on the required operability of the Residual Heat Removal

Service Water (RHRSW) system. As required by TS Section 3.7.1.1, the

RHRSW subsystem(s) associated with the required operable RHR heat

exchanger(s) will continue to remain operable. Each operable RHRSW

subsystem consists of two (2) operable pumps and the required

operable flowpath to provide decay heat removal via the associated

RHR heat exchanger.

The RHRSW system piping is designed, fabricated, inspected, and

tested in accordance with the requirements of ASME [American Society

of Mechanical Engineers], Section III Class 3, and each RHRSW

subsystem is single active failure proof in that the failure of a

motor-operated valve, diesel generator, or pump does not prevent the

system from performing its safety function.

The required availability of four (4) loops of the Low Pressure

Coolant Injection (LPCI) mode of RHR during OPCONs 1, 2, and 3 as

required by TS Section 3.5.1 is not impacted by the four (4)

``subsystem'' Shutdown Cooling designation. No change to any RHR

system instrumentation logic, required Emergency Core Cooling System

(ECCS) availability, or method of operation is involved.

NUREG-1433, Revision 1, also re-affirms that each Shutdown

Cooling ``subsystem'' is considered operable if it can be manually

aligned, remotely or locally, in the shutdown cooling mode for

removal of decay heat. Thus, a LPCI-dedicated pump can be aligned

for LPCI automatic initiation, yet still be considered part of an

operable shutdown cooling subsystem as long as it can be re-aligned

for Shutdown Cooling.

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

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed TS changes do not involve any physical changes to

plant structures, systems, or components. The RHR system will

continue to function as designed in all modes of operation. No new

accident type is created as a result of the proposed changes. No new

failure mode for any equipment is created. The changes are

consistent with the guidance provided in NUREG-1433, Revision 1,

pertaining to RHR Shutdown Cooling operation in OPCONs 4 and 5.

The four (4) ``subsystem'' Shutdown Cooling designation has no

effect on the required operability of the RHRSW system. The RHRSW

subsystem(s) associated with the required operable RHR heat

exchanger(s) will continue to remain operable as required by TS

Section 3.7.1.1. Each operable RHRSW subsystem consists of two (2)

operable pumps and the required operable flowpath to provide decay

heat removal via the associated RHR heat exchanger.

The RHRSW system piping is designed, fabricated, inspected, and

tested in accordance with the requirements of ASME, Section III,

Class 3, and each RHRSW subsystem is single active failure proof in

that the failure of a motor-operated valve, diesel generator, or

pump does not prevent the system from performing its safety

function.

The required availability of four (4) loops of the LPCI mode of

RHR during OPCONs 1, 2, and 3 as required by TS Section 3.5.1 and

3.5.2 is not impacted by the four (4) ``subsystem'' Shutdown Cooling

designation. No change to any RHR system instrumentation logic,

required ECCS availability, or method of operation is involved.

NUREG-1433, Revision 1, also re-affirms that each Shutdown

Cooling ``subsystem'' is considered operable if it can be manually

aligned, remotely or locally, in the Shutdown Cooling mode for

removal of decay heat. Thus, a LPCI-dedicated pump can aligned be

[sic] [be aligned] for automatic LPCI initiation, yet still be

considered part of an operable shutdown cooling subsystem as long as

it can be re-aligned for Shutdown Cooling.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

Although the Bases for TS Sections 3/4.4.9.2, 3/4.9.11.1, and 3/

4.9.11.2 are being revised in support of this proposed TS change,

the changes only involve providing clarification regarding the

designation of the RHR Shutdown Cooling operation configuration in

OPCONs 4 and 5. The proposed TS changes do not involve any physical

changes to plant structures, systems, or components. The RHR system

will continue to function as designed in all modes of operation. The

consequences of equipment malfunction are not changed from those in

existing analyses, with no increase in onsite or offsite

radiological effects. The RHR system will continue to function as

designed to mitigate the consequences of an accident and resultant

onsite and offsite radiological effects remain as previously

evaluated. The proposed changes are consistent with the guidance

stipulated in NUREG-1433, Revision 1.

The four (4) ``subsystem'' Shutdown Cooling designation has no

effect on the required operability of the RHRSW system. As required

by TS 3.7.1.1, the RHRSW subsystem(s) associated with the required

operable RHR heat exchanger(s) will continue to remain operable.

Each operable RHRSW subsystem consists of two (2) operable pumps and

the required operable flowpath to provide decay heat removal via the

associated RHR heat exchanger.

The RHRSW system piping is designed, fabricated, inspected, and

tested in accordance with the requirements of ASME, Section III,

Class 3, and each RHRSW subsystem is single active failure proof in

that the failure of a motor-operated valve, diesel generator, or

pump does not prevent the system from performing its safety

function. (In the same manner that manual action may be required for

RHR system alignment in OPCONs 4 and 5 with one (1) RHR heat

exchanger operable, a failure of the motor-operated RHRSW inlet or

outlet heat exchanger isolation valves may require manual

positioning for the required alignment.)

The required availability of four (4) loops of the LPCI mode of

RHR during OPCONs 1, 2, and 3* as required by TS Section 3.5.1 is

not affected by the four (4) ``subsystem'' Shutdown Cooling

configuration. No change to any RHR system instrumentation logic,

required ECCS availability, or method of operation is involved.

NUREG-1433, Revision 1, also re-affirms that each Shutdown

Cooling ``subsystem'' is

[[Page 55038]]

considered operable if it can be manually aligned, remotely or

locally, in the Shutdown Cooling mode for removal of decay heat.

Thus, a LPCI-dedicated pump can be aligned for LPCI automatic

initiation, yet still be considered part of an operable Shutdown

Cooling ``subsystem'' as long as it can be re-aligned for Shutdown

Cooling.

Therefore, the proposed TS changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

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

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

Pennsylvania

Date of amendment request: June 28, 1996

Description of amendment request: The proposed Technical

Specifications (TS) changes would incorporate performance-based

testing, in accordance with 10 CFR Part 50, Appendix J, ``Primary

Reactor Containment Leakage Testing For Water-Cooled Power Reactors,''

Option B. This option allows utilities to extend the frequencies of the

Type A Containment (ILRT) Leak Rate Test and Type B and C Local Leak

Rate Tests (LLRTs) based on the performance and design of the

containment and components.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

in the probability or consequences of an accident previously

evaluated.

Incorporation of the new 10 CFR 50, Appendix J, Option B at LGS,

Units 1 and 2 does not increase the probability of occurrence of an

accident previously evaluated. The containment structure including

its isolation capability is not an accident initiator.

These changes do not involve any changes to the containment

structure, system or components which could increase the probability

of occurrence of an accident previously evaluated or act as a new

accident initiator. Implementation of the proposed changes will

affect the manner in which these structures, systems, or components

(SSCs) are tested; however, the new testing schedule is not an

initiator of any analyzed event. No equipment changes are involved

with adoption of Option B; therefore, performance-based test

intervals for Type A, B, and C tests do not increase the probability

of occurrence of a malfunction of equipment important to safety

previously evaluated. No physical changes are being made to the

plant, nor are there any changes being made in the operation of the

plant as the result of increasing the test intervals. Additionally,

the proposed TS changes will not alter the operation of equipment

available for the mitigation of accidents or transients, therefore,

this change will not result in any significant increase to onsite or

offsite dose previously evaluated. The potential for time-based and

activity-based failure mechanisms which could lead to excessive

containment leakage has been determined to be minimal. Performance-

based test intervals for Type A, B, and C tests will not alter any

safety limits which ensure the integrity of fuel barriers, and will

not increase the primary containment leakage limits.

Performance-based test intervals for Type A, B, and C leak tests

do not increase the consequences of an accident previously

evaluated. NUREG-1493 concluded that reducing the frequency of Type

A tests from the current three per ten years to one per ten years

was found to lead to an imperceptible increase in risk. NUREG-1493

includes the results of a sensitivity study performed to explore the

risk impact of several alternative leak rate test schedules. The

estimated increase in population exposure risk ranged from 0.02% to

0.14%. The risk impact was determined to be very small since Type B

and C testing (local leak rate tests) detect a very large percentage

of overall containment leakages. The percentage of leakages detected

by Type A tests is very small. Past test results experienced at

Limerick Units 1 and 2 concur with these determinations. NUREG-1493

also concluded that the overall unit risk is not very sensitive to

changes in containment leakage rates. Given the insensitivity of

risk to containment leak rates and the small fraction of leak paths

detected solely by the Type A tests, increasing the interval between

Type A tests is possible with minimal impact on public risk.

NUREG-1493 also concluded that, based on a model of component

failure with time, the performance-based alternatives to current,

local-leakage testing requirements are feasible without significant

risk impact. The LGS design and past performance is bounded by the

NUREG study. The NUREG model indicated that the number of components

tested could be reduced by about 60% with less than a three-fold

increase in the incremental risk due to containment leakage. Since

under existing requirements, leakage contributes less than 0.1

percent of overall accident risk, the overall impact is very small.

Therefore, the proposed TS changes will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Performance-based test intervals for Type A, B, and C leak tests

do not introduce a new or different type of accident or create the

possibility of a different type of malfunction of equipment

important to safety than previously evaluated. No physical changes

are being made to the plant, nor are there any changes being made in

the operation of the plant as the result of increasing the test

intervals. No new failure modes of plant equipment previously

evaluated will be introduced. Additionally, the TS changes will not

alter the operation of equipment available for the mitigation of

accidents or transients. The safety function of the primary

containment will be retained since the containment will continue to

provide an essentially leak tight barrier against the uncontrolled

release of radioactivity to the environment for postulated accidents

previously evaluated.

Therefore, the proposed TS changes will not create the

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

previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The margin of safety is not reduced as a result of adopting 10

CFR 50, Appendix J, Option B. The effect of increasing containment

leakage rate testing intervals was evaluated in NUREG-1493 using

historical industry leakage rate testing results. Performance

history at LGS is consistent with the conclusions reached in NUREG-

1493 and NEI 94-01. The results of the NUREG evaluation conclude

that the increased safety risk corresponding to the extended test

intervals is small (less than 0.1% of total risk). The revised TS

will continue to maintain the allowable leakage rate for the Type A

tests. In addition, the requirement to perform a periodic general

visual inspection of the primary containment has been maintained at

the original interval of three times in 10 years as part of the

performance-based leakage rate testing program.

The risk of a non-detectable increase of primary containment

leakage is considered to be negligible due to the conclusion that 10

CFR 50, Appendix J, Type B and C testing program will continue to be

conducted between Type A tests. A review of previous LGS Type A test

results has concluded that the only failure mechanisms are activity-

based. There is no indication of time-based failures that would not

be identified during the performance of Type B and C tests.

Therefore, we have concluded that the proposed adoption of the

Option B intervals would not result in a non-detectable primary

containment leakage rate in excess of the allowable value (i.e.,

0.5% wt/day) established by the LGS TS.

The proposed TS will continue to maintain the allowable leakage

rate for the combined Type B and C tests. As supported by the

findings of NUREG-1493, the percentage of leakages detected by Type

A tests is small (as

[[Page 55039]]

stated above) and Type B and C leakage tests are capable of

detecting more than 97% of containment leakages and virtually all

such leakages are identified by local leak rate tests of containment

isolation valves. The Type B and C test intervals will be

established through the PCLRTP for each component based on design

and previous LGS test performance history.

Therefore, the proposed TS changes do not involve a reduction in

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

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: September 25, 1996

Description of amendment request: The amendments would relocate to

the Salem Updated Final Safety Analysis Report the list of containment

isolation valves that are currently located in Table 3.6-1 of Technical

Specification 3.6.3. In addition, references to the table in

specifications 1.7, 3.6.1, and 3.6.3 are being updated.

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 consequence of an accident previously

evaluated.

The proposed changes simplify the TS, meet the regulatory

requirements for control of containment isolation, and are

consistent with the guidance provided in Generic Letter (GL) 91-08,

``Removal of Component Lists from Technical Specifications.'' The

procedural details of TS Table 3.6-1 have not been changed, only

relocated to a different controlling document, the Salem Update

[sic] [Updated] Final Safety Analysis Report (UFSAR). The proposed

changes are administrative in nature, should result in improved

administrative practices, and do not affect plant operations.

The probability of occurrence of a previously evaluated accident

is not increased because this change does not introduce any new

potential accident initiating conditions. The consequences of an

accident previously evaluated is not increased because the ability

of containment to restrict the release of any fission product

radioactivity to the environment will not be degraded by this

change.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes are administrative in nature, do not result

in a physical alterations or changes to the operation of the plant,

and cause no change in the method by which any safety-related system

performs its functions. Therefore, this proposed change will not

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

any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The administrative change to relocate TS Table 3.6-1 to the

UFSAR does not alter the basic regulatory requirements for

containment isolation and will not adversely affect the containment

isolation capability for credible accident scenarios. Adequate

control of the content of the relocated table is assured by the

10CFR50.59 review process.

The proposed relocation of TS Table 3.6-1 does not alter the

requirements for CIV operability currently in the TS. the Limiting

Condition for Operation and the Surveillance Requirements would be

retained in the revised TS. Therefore, the proposed changes will not

affect the meaning, application, and function of the current TS

requirements for the CIVs.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079

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

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

NRC Project Director: John F. Stolz

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: September 25, 1996

Description of amendment request: The amendments would change

Technical Specification 3/4.8.1, ``Electrical Power Systems,'' to

revise the Emergency Diesel Generator (EDG) voltage and frequency

limits as a result of updated EDG load calculations and to eliminate

ambiguity in the testing methodology for EDG start timing.

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.

Since no change is being made to the offsite power supplies, or

to any system or component that interfaces with the offsite power

supplies, there is no change in the probability of a Loss of Offsite

Power Accident.

The proposed changes provide the necessary conservatism for

voltage and frequency to ensure the EDGs are not run in an

overloaded condition and that driven equipment is not damaged during

steady state operation following a Loss of Offsite Power coincident

with a Loss of Coolant Accident. Since the narrower band of voltage

and frequency for the isochronous mode continues to ensure proper

steady state operation of the EDG and associated driven equipment,

there is no change in the consequences of an accident previously

evaluated.

Based on the above, the proposed amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed amendment does not result in any design or physical

configuration changes to the EDGs. Proposed changes made to the

testing parameters and testing methodology will not cause a new or

different accident since the EDGs are used for accident mitigation

and no new failure modes are being introduced. Therefore, the

proposed amendment will not create the possibility of a new or

different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed amendment provides further conservatism to the

voltage and frequency band currently specified in the TSs. The

proposed voltage and frequency changes ensure the EDG will not be

overloaded from an over-frequency condition and driven equipment

will not be damaged from an over-voltage condition.

The control system is set to control the EDG voltage within the

bands specified in the requested changes. The changes are consistent

with current calculations and within the capability of the controls.

Since the narrower band of voltage and frequency for the isochronous

mode is bounded by the existing TS, there is no change in the margin

of safety. The increased band for droop mode will ensure the EDG is

capable of operating in accordance with normal offsite power

parameters and does not reduce 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

[[Page 55040]]

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079

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

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

NRC Project Director: John F. Stolz

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: October 1, 1996

Description of amendment request: The proposed amendments would

change Technical Specifications (TSs) 3/4.7.1.5, ``Main Steam Line

Isolation Valves (MSIVs),'' and 3/4.3.2, ``Engineered Safety Feature

Actuation System Instrumentation.'' These changes are needed to

accommodate entry into Modes 3 and 2 prior to performing MSIV closure

time testing in Mode 2. The proposed amendments would also allow for

the repair and testing of inoperable MSIVs in certain operating Modes,

and would change the low steam line pressure trip setpoint value for

safety injection to make it consistent with the previously approved

value for steam line isolation.

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 isolation capability of the MSIVs and the protective

functions of the low steam line pressure channels are necessary for

accident mitigation and do not impact the probability of an

accident. MSIV testing in the higher modes is necessary to obtain

conditions which enable testing of the MSIVs. These conditions are

consistent with the current accident analyses for main steam line

breaks and secondary system depressurization. Failure of a MSIV,

which could be encountered during testing, is accounted for in the

accident analyses.

Provisions for entering Mode 2 within six hours with an

inoperable MSIV allows operators to remove the plant from power

generation in a more controlled manner without challenging plant

safety systems and is consistent with other plant shutdown TS (i.e.,

TS 3.0.3). The additional six hours to Hot Shutdown, should MSIV

closure be infeasible, does not result in a significant increase in

the probability or consequence of an accident since this is a very

small incremental time addition. The values for the low steam line

pressure safety injection are higher and are bounded by the present

accident analysis. The elimination of the obsolete stroke time of

eight seconds is editorial in nature. As a result, the changes

proposed do not involve a significant increase in the probability or

consequence of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not involve any modifications to

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

systems, do not introduce any new operating configurations or new

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

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

new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

MSIV testing in Mode 2 is within the currently analyzed plant

operation as discussed in the Updated Final Safety Analysis Report

(UFSAR) Sections 10.3 and 15.4. These UFSAR sections address

performance of the TS surveillance test at or near 1000 psig Steam

Generator pressure to assure main steam isolation occurs within the

accident conditions, where Steam Generator pressure may be lower

during Mode 1 operation. The test methodology demonstrating MSIV

operability is consistent with the accident analysis.

Operation in Modes 2 and 3 with one or more isolation valve

inoperable and in the closed position does not impact the margin of

safety since the valves are already performing the safety function.

The protective functions that occur as a result of the low steam

line pressure initiating signal remain bounded by the values assumed

in the safety analyses. That is, the protective functions that occur

as a result of this initiating signal already assume a setpoint that

is conservative for the revised value. The change to the setpoint

eliminates conflicting information in the TS.

Therefore, 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 10 CFR 50.92(c) are

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079

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

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

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-311, Salem

Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

Date of amendment request: September 20, 1996, as supplemented

September 30, 1996

Description of amendment request: The proposed amendment would

change Technical Specification 4.7.7.b.4 to indicate that the specified

flowrate for the Auxiliary Building Exhaust Air Filtration System

applies only to system testing.

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 accident considered in this proposed change is the Loss of

Coolant Accident (LOCA) as described in Section 15.4 of the UFSAR

[Updated Final Safety Analysis Report]. The assumption is that:

``The Auxiliary Building Ventilation System will discharge the vapor

(from recirculation liquid leakage) to the atmosphere through

charcoal filters which have an efficiency of 90 percent.'' As such

the system acts to limit the total offsite and control room

radiation doses following a LOCA.

The Auxiliary Building Ventilation System [ABVS] is designed to

maintain the Auxiliary Building at a negative pressure with respect

to the atmosphere during normal and emergency operation. Filtration

of radio-iodines is accomplished by administratively aligning the

ECCS [emergency core cooling system] equipment areas exhaust flows

to the standby charcoal adsorber bed if required. The ABVS has no

direct impact on reactor operation or on any system connected to the

Reactor Coolant Pressure Boundary.

The emergency operation of the Auxiliary Building Ventilation

System is not affected by the proposed changes. The acceptance

criteria for system performance are not modified by the requested

change. The change clarifies the intent of SR [surveillance

requirement] 4.7.7.b.4 and the basis for the flowrates used for

system acceptance testing. It has been determined that operation of

the system at lower flow rates than those specified for surveillance

testing is conservative with respect to the radio-iodine removal

efficiency assumed for the charcoal adsorber. A higher removal

efficiency results in lower total exposures at the site boundary and

within the control room. Additionally, the system is capable of

maintaining the required negative pressure at the reduced flowrate.

Given the above, it is concluded that the proposed change does

not result in an increase in the probability or consequences

associated with previously analyzed accidents.

[[Page 55041]]

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

or different kind of accident from any accident previously

evaluated.

The proposed amendment does not result in any design or

operational change to the ABVS, to the Nuclear Steam Supply System,

to the ECCS System, to the Containment Building, to the fuel or to

the electrical power supplies. Therefore, the proposed amendment

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

accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Specification 3/4.7.7 and the associated bases were reviewed to

determine if the proposed changes result in a reduction in the

margin of safety. The change to SR 4.7.7.b.4 continues to assure

that the system is operated consistent with the assumptions of the

accident analysis. The proposed changes to Bases 3/4.7.7 clarify the

basis for flowrates associated with ABVS surveillance test

requirements. All changes result in ABVS operation that is just as

conservative as that assumed in existing analyses.

The proposed changes do not involve the addition or modification

of plant equipment, are consistent with the design basis of the ABVS

as described in the UFSAR, and appropriately limit operation to be

consistent with the assumptions of the accident analysis. As such

there is no reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079

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

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

NRC Project Director: John F. Stolz

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.

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

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

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

Date of application for amendments: June 17, 1996

Brief description of amendments request: The proposed amendments

would modify the technical specifications to change (1) the reference

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

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

pressurizer pressure to account for new instrument uncertainties and to

reduce the allowed operating band.

Date of individual notice in Federal Register: September 11, 1996

(61 FR 47963)

Expiration date of individual notice: October 11, 1996

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

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

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

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

Date of application for amendments: June 28, 1996

Brief description of amendments request: The proposed amendments

would modify the technical specifications to increase the minimum

required amount of anhydrous trisodium phosphate (TSP) in the

containment baskets.

Date of individual notice in Federal Register: September 11, 1996

(61 FR 47962), as corrected September 26, 1996 (61 FR 50535).

Expiration date of individual notice: October 11, 1996

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

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

Station, Unit No. 1, Washington County, Nebraska

Date of application for amendment: August 23, 1996

Brief description of amendment request: The proposed amendment

would revise Paragraph 2.B(2) of

Facility Operating License No. DPR-40 to allow source materials in

the form of depleted or natural uranium as reactor fuel and to revise

Technical Specification 4.3.2 to include depleted uranium in describing

the reactor core.

Date of individual notice in Federal Register: August 30, 1996 (61

FR 45995)

Expiration date of individual notice: September 30, 1996

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

South 15th Street, Omaha, Nebraska 68102

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,

Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two

Creeks, Manitowoc County, Wisconsin

Date of application for amendment: September 19, 1996

Brief description of amendment request: The proposed amendments

would change Technical Specification requirements related to the low

temperature overpressure protection (LTOP) system. Specifically, the

reactor coolant system (RCS) temperature below which LTOP is required

to be enabled and one high pressure safety injection pump is required

to be rendered inoperable would be changed from 275 deg.F to 355

deg.F. Also, a specification would be added stating that only one

reactor coolant pump shall be operated when the RCS temperature is less

than or equal to 125 deg.F. Finally, editorial changes would be made

to rename the ``Overpressure Mitigating System'' as the ``Low

Temperature Overpressure Protection System.'' Date of individual notice

in Federal Register: October 1, 1996 (61 FR 51308) Expiration date of

individual notice: October 31, 1996

Local Public Document Room location: Joseph P. Mann Library, 1516

Sixteenth, Two Rivers, Wisconsin 54241

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: September 27, 1996

Brief description of amendment request: The proposed amendment

would change Technical Specification (TS) requirements related to the

low temperature overpressure protection (LTOP) system. Specifically,

the LTOP curve would be modified to define 10 CFR Part 50, Appendix G

pressure temperature limitations for LTOP evaluation through the end of

operating cycle (EOC) 33. In addition, the LTOP enabling temperature

and the temperature required for starting a reactor coolant pump would

be changed consistent with the design basis for the LTOP system.

Finally, the TS bases would be changed consistent with he changes

described above.

Date of individual notice in Federal Register: October 7, 1996 (61

FR 52472)

[[Page 55042]]

Expiration date of individual notice: November 6, 1996

Local Public Document Room location: University of Wisconsin,

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

Notice Of Issuance Of Amendments ToFacility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

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

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

the particular facilities involved.

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

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

North Carolina

Date of application for amendment: July 19, 1996

Brief description of amendment: The amendment revises the

containment spray nozzle surveillance interval in TS 3/4.6.2 from 5 to

10 years.

Date of issuance: October 3, 1996

Effective date: October 3, 1996

Amendment No.: 67

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

Technical Specifications.

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

44354) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 3, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

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

Date of amendment request: April 11, 1996, as supplemented August

23, 1996

Brief description of amendments: The amendments revised the

Technical Specifications to permit implementation of 10 CFR Part 50,

Appendix J, Option B.

Date of issuance: October 3, 1996

Effective date: October 3, 1996

Amendment Nos.: 185 and 176

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

revised the Technical Specifications.

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

20846) The additional information contained in the supplemental letter

dated August 23, 1996, was clarifying in nature and thus, within the

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

significant hazards consideration determination.The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated October 3, 1996.No significant hazards consideration

comments received: No.

Public Document Room location: Tomlinson Library, Arkansas Tech

University, Russellville, AR 72801

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

Unit No. 1, Pope County, Arkansas

Date of amendment request: April 29, 1996

Brief description of amendment: The amendment relocated cycle

specific operating parameters from the Technical Specifications to the

Core Operating Limits Report per Generic Letter 88-16. The parameters

being relocated by this amendment include the variable low reactor

coolant system pressure trip and the variable low reactor coolant

system pressure-temperature protective limits.

Date of issuance: October 3, 1996

Effective date: October 3, 1996

Amendment No.: 186

Facility Operating License No. DPR-51: Amendment revised the

Technical Specifications.

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

28613) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 3, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

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

Date of amendment request: November 7, 1995, as supplemented by

letter dated April 11, 1996.

Brief description of amendment: The amendment modifies the Appendix

A Technical Specifications related to Safety Injection Tank level and

pressure setpoints.

Date of issuance: September 27, 1996

Effective date: September 27, 1996

Amendment No.: 121

Facility Operating License No. NPF-38: Amendment revised the

Technical Specifications.

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

FR 58401) The additional information contained in the supplemental

letter dated April 11, 1996, was clarifying in nature and thus, within

the scope of the initial notice and did not affect the staff's proposed

no significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated September 27, 1996.No significant hazards consideration comments

received: No.

Local Public Document Room location: University of New Orleans

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

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: July 17, 1996

Brief description of amendments: The amendments consist of changes

to the Technical Specifications regarding containment leakage tests.

Date of issuance: October 4, 1996

Effective date: October 4, 1996

Amendment Nos.: 192 and 186Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

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

44357)

[[Page 55043]]

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

Safety Evaluation dated October 4, 1996.No significant hazards

consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Georgia Power Company, Oglethorpe Power Corporation, Municipal

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

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

Appling County, Georgia

Date of application for amendments: May 21, 1996

Brief description of amendments: The amendments revise the

condensate storage tank level indication to ensure that the water level

is sufficient to provide 50,000 gallons of water for core spray makeup

to the reactor pressure vessel. On September 24, 1996, based on a

teleconference between the licensee and the NRC project manager, it was

mutually agreed to change the requested implementation schedule from 90

days to 30 days.

Date of issuance: October 2, 1996

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

30 days

Amendment Nos.: 202 and 143

Facility Operating License Nos. DPR-57 and NPF-5: Amendments

revised the Technical Specifications.

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

44358) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 2, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island

Nuclear Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania

Date of application for amendment: January 16, 1995

Brief description of amendment: This amendment revised the

Technical Specification to incorporate an improvement from

administrative controls section of the revised standard TS for B&W

plants.

Date of issuance: October 8, 1996

Effective date: October 8, 1996

Amendment No.: 50Possession-Only License No. DPR-73: The amendment

revised the Technical Specifications.

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

FR 65679). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 8, 1996No significant

hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,

Linn County, Iowa

Date of application for amendment: July 5, 1996

Brief description of amendment: The amendment will support the

implementation of noble metal chemical addition at the Duane Arnold

Energy Center as a method to enhance the effectiveness of hydrogen

water chemistry in mitigating intergranular stress corrosion cracking

in reactor vessel internal components. Specifically, the amendment will

permit an increase of the reactor water conductivity limit in Technical

Specification (TS) Table 3.6.B.2-1 and several other changes in TS

sections 4.6.B.2.c, 4.6.B.2.d, and the associated Bases.

Date of issuance: October 3, 1996

Effective date: October 3, 1996

Amendment No.: 218

Facility Operating License No. DPR-49: Amendment revised the

Technical Specifications.

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

40020) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 3, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids, Iowa 52401

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,

Linn County, Iowa

Date of application for amendment: December 22, 1995, as

supplemented September 20, 1996

Brief description of amendment: The amendment revises the Duane

Arnold Energy Center (DAEC) Technical Specifications (TS) Sections

3.7.A and 4.7.A, ``Primary Containment,'' by deleting information also

contained in 10 CFR Part 50, Appendix J, Option A and incorporating

references to the Primary Containment Leakage Rate Testing Program.

These changes allow the use of the performance based option of

containment leak testing. The amendment also adds Operability and

Surveillance Requirements (SRs) for the drywell air lock. Minor

administrative changes were also made. These changes are consistent

with comparable specifications in the Improved Standard Technical

Specifications (ITS), NUREG-1433. In addition, the staff executed

administrative changes and corrections to the TS Bases, as submitted in

two letters dated February 13, 1995. Sections changed or corrected are

Section 1.2, Bases; Section 2.2, Bases Reactor Coolant System

Integrity; Section 3.7.H/4.7.H, Bases Containment Atmosphere Dilution;

and Section 3.7.I/4.7.I, Bases Oxygen Concentration.

Date of issuance: October 4, 1996

Effective date: October 4, 1996

Amendment No.: 219

Facility Operating License No. DPR-49: Amendment revised the

Technical Specifications.

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

3499) The September 20, 1996, submittal was clarifying in nature and

did not affect the no significant hazards determination. The

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

Safety Evaluation dated October 4, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids, Iowa 52401

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

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

Illinois

Date of application for amendment: June 28, 1996 and as

supplemented on September 17, 1996

Brief description of amendment: The amendment will allow removal of

the Inclined Fuel Transfer System (IFTS) primary containment blind

flange while primary containment is required to be operable. This will

provide flexibility to operate the IFTS for the purpose of testing and

exercising the system during such conditions. Primary containment

integrity will be provided by an alternate means while the blind flange

is removed. The change will be incorporated via a provisional note into

Technical Specification (TS) Surveillance Requirement 3.6.1.3.3,

associated with TS 3.6.1.3, ``Primary Containment Isolation Valves

(PCIVs).''

Date of issuance: October 3, 1996

Effective date: October 3, 1996

Amendment No.: 107

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

Technical Specifications.

[[Page 55044]]

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

40021) The information provided in the licensee's letter of September

17, 1996 provided clarifying information and did not involve

significant changes to the original Federal Register notice.The

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

Safety Evaluation dated October 3, 1996.No significant hazards

consideration comments received: No

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

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

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

Illinois

Date of application for amendment: June 21, 1996, and as

supplemented by letter dated August 15, 1996

Brief description of amendment: The amendment modifies Section 5.7,

``High Radiation Areas,'' of the ``Administrative Controls'' section of

the Clinton Power Station technical specifications (TS). The changes

include: (1) allowing utilization of a Radiation Work Permit (RWP) ``or

equivalent'' to control entry into a high radiation area; (2)

clarifying the example given in the TS of individuals who are qualified

in radiation protection procedures; (3) clarifying the requirements for

when specified access controls and barriers for high radiation areas

within large areas like the containment may be established; (4)

clarifying that it is acceptable for an RWP to specify a maximum dose,

i.e., a specified setpoint on an alarming dosimeter in lieu of a stay

time for entry into a high radiation area (where an individual could

receive a deep dose equivalent of 3000 mrem in one hour); (5)

eliminating the upper dose limit for specifying the applicability of

the requirements of Specification 5.7.1; (6) providing additional

flexibility regarding the control of keys to locked doors for

preventing unauthorized entry into high radiation areas; (7) providing

alternate means of informing individuals of dose rates in immediate

work areas; (8) reorganizing TS Sections 5.7.1, 5.7.2, and 5.7.3 into

four sections (5.7.1, 5.7.2, 5.7.3 and 5.7.4); and (9) making minor

edits to enhance readability.

Date of issuance: October 3, 1996

Effective date: October 3, 1996

Amendment No.: 108

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

Technical Specifications.

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

40021) The August 21, 1996, submittal consisted of supporting technical

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

significant hazards consideration determination or expand the scope of

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

amendment is contained in a Safety Evaluation dated October 3, 1996.No

significant hazards consideration comments received: No.

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

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

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

Date of application for amendment: May 2, 1996, as supplemented by

letter dated August 30, 1996

Brief description of amendment: The amendment removes Technical

Specification Figure 5.1, which was used in maintaining Keff

values, and substitutes in its place a defined requirement for maximum

Kinfinity for any fuel placed in the Millstone Unit 1 spent fuel

pool.

Date of Issuance: October 4, 1996

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

within 60 days.

Amendment No.: 97

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

Technical Specifications.

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

37301) The August 30, 1996, letter provided additional, clarifying

information that did not change the scope of the May 2, 1996,

application and the initial proposed no significant hazards

consideration determination.The Commission's related evaluation of this

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

significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, Connecticut

Northern States Power Company, Docket No. 50-282, Prairie Island

Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota

Date of application for amendment: July 15, 1996, and supplemented

August 22, 1996

Brief description of amendment: The amendment allows the use of the

moveable in-core detector system for measurement of the core peaking

factors with less than 75 percent and greater than or equal to 50

percent of the detector thimbles available. The amendment is a one-time

only change for Prairie Island, Unit 1, to reduce the number of

required in-core detectors necessary for continued operation for the

remainder of Operating Cycle 18.

Date of issuance: October 10, 1996

Effective date: October 10, 1996, and shall remain effective for

the remainder of Cycle 18 only

Amendment No.: 124

Facility Operating License No. DPR-42. Amendment revised the

Technical Specifications.

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

40024) By letter dated August 22, 1996, NSP forwarded a copy of the

results of its most recent low power physics tests to the NRC for use

as a reference and provided additional clarifying information. This

information was within the scope of the original application and did

not change the staff's initial proposed no significant hazards

considerations determination. Therefore, renoticing was not

warranted.The Commission's related evaluation of the amendments is

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

hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: May 17, 1996

Brief description of amendment: The amendment revises Technical

Specifications (TS) 2.18, 3.14, 3.3, and 5.10 to relocate the

operability requirements for shock suppressors (snubbers) from the TS

to the Updated Safety Analysis Report (USAR) and incorporate snubber

examination and testing requirements in TS 3.3.

Date of issuance: September 27, 1996

Effective date: September 27, 1996

Amendment No.: 176

[[Page 55045]]

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

Technical Specifications.

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

44360) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 27, 1996.No

significant hazards consideration comments received: No.

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

South 15th Street, Omaha, Nebraska 68102

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

Station,Unit No. 1, Washington County, Nebraska

Date of amendment request: August 23, 1996

Brief description of amendment: The amendment modifies paragraph

2.B.(2) of

Facility Operating License No. DPR-40 allowing the use of source

material, in the form of depleted or natural uranium, as reactor fuel.

Date of issuance: October 2, 1996

Effective date: October 2, 1996

Amendment No.: 177

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

Operating License.

Date of initial notice in Federal Register: August 30, 1996 (61 FR

45995) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 2, 1996.No significant

hazards consideration comments received: No.

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

South 15th Street, Omaha, Nebraska 68102

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

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: January 25, 1996

Brief description of amendment: The amendment would extend the

instrumentation surveillance test intervals to support 24-month

operating cycles. These proposed changes would eliminate the mid-cycle

outages to perform the Technical Specification surveillance

requirements.

Date of issuance: October 2, 1996

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

30 days.

Amendment No.: 233

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

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

25709) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 2, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

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

York 13126

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

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: March 27, 1996, as supplemented

April 24, 1996, and August 15, 1996

Brief description of amendment: The proposed amendment changes

would permit implementation of 10 CFR Part 50, Appendix J, Option B

with an exception to the guidelines of Regulatory Guide 1.163 for Type

C testing of primary containment isolation valves in the reverse (non-

accident) direction.

Date of issuance: October 4, 1996

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

30 days.

Amendment No.: 234

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

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

20855) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

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

York 13126

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

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: August 9, 1996, as supplemented

September 17, 1996

Brief description of amendment: The amendment revises the Technical

Specifications to revise the safety limit minimum critical power ratio

for cycle 19 operation from its current value of 1.07 (for the fuel

currently in the reactor for cycle 18) for two recirculation loop

operation to 1.10, and from 1.08 to 1.12 for single recirculation loop

operation.

Date of issuance: October 4, 1996

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

within 30 days.

Amendment No.: 150

Facility Operating License No. DPR-28: Amendment revised the

Technical Specifications.

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

44364) The September 17, 1996, letter provided clarifying information

that did not change the scope of the August 9, 1996, application and

initial proposed no significant hazards consideration determination.The

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

Safety Evaluation dated October 4, 1996.No significant hazards

consideration comments received: No

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

South Carolina Electric & Gas Company, South Carolina Public

Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear

Station, Unit No. 1, Fairfield County, South Carolina

Date of application for amendment: April 16, 1996, as supplemented

July 25, 1996

Brief description of amendment: The amendment permits

implementation of 10 CFR Part 50, Appendix J, Option B, ``Performance-

Based Requirements.''

Date of issuance: October 2, 1996

Effective date: October 2, 1996

Amendment No.: 135

Facility Operating License No. NPF-12: Amendment revises the

Technical Specifications.

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

34898) The July 25, 1996, supplement provides clarifying information

and did not change the scope of the initial notice. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated October 2, 1996.No significant hazards consideration comments

received: No

Local Public Document Room location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180

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

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

California

Date of application for amendment: March 13, 1996

Brief description of amendment: The change revises the San Onofre

Unit 1 License Condition 2.D. This change eliminates a reporting

requirement that is redundant to reporting requirements in 10 CFR 50.72

and 50.73. Additionally, the amendment makes administrative and

editorial changes to the Permanently Defueled Technical Specifications.

Date of issuance: October 3, 1996

[[Page 55046]]

Effective date: October 3, 1996

Amendment No.: 158

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

Technical Specifications.

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

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

contained in a Safety Evaluation dated October 3, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Science Library, University of

California, Irvine, California 92713

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: December 6, 1995, as

supplemented by letters dated August 30, 1996, and September 20, 1996

Brief description of amendments: These amendments revise Technical

Specifications (TS) Section 4.3 ``Fuel Storage'' to allow fuel

assemblies having a maximum U-235 enrichment of 4.8 weight percent (w/

o) to be stored in both the spent fuel racks and the new fuel racks.

Additionally, TS Section 3.7.18 ``Spent Fuel Assembly Storage,''

Figures 3.7.18-1 ``Unit 1 Fuel Minimum Burnup vs. Initial Enrichment

for Region II Racks,'' and 3.7.18-2 ``Units 2 and 3 Fuel Minimum Burnup

vs. Initial Enrichment for Region II Racks,'' are being revised and

relabeled.

Date of issuance: October 3, 1996

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

as of the date of issuance.

Amendment Nos.: Unit 2 - 131; Unit 3 - 120

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

revised the Technical Specifications.

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

15997) The August 30, 1996, and September 20, 1996, letters provided

additional clarifying information and did not change the initial no

significant hazards consideration determination.The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated October 3, 1996.No significant hazards consideration

comments received: No.

Temporary Local Public Document Room location: Science Library,

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

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

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

Date of amendment request: July 31, 1996 (TXX-96433)

Brief description of amendments: The amendments revised core safety

limit curves (Technical Specification (TS) Figure 2.1-1a) and new N-16

setpoint values and parameters (TS Table 2.1-1) for Unit 1, and

reference to topical report RXE-95-001-P as an approved methodology for

small break loss of coolant accident analysis for Units 1 and 2.

Date of issuance: September 30, 1996

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

days

Amendment Nos.: 52 and 38

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

revised the Technical Specifications.

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

44362) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 30, 1996.No

significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

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

19497, Arlington, TX 76019

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

Callaway County, Missouri

Date of application for amendment: April 12, 1996, as supplemented

by letters dated August 2, 1996, August 19, 1996, and September 5,

1996.

Brief description of amendment: The amendment revises the Technical

Specifications to address the installation of laser welded tube sleeves

in the Callaway Plant steam generators.

Date of issuance: October 1, 1996

Effective date: October 1, 1996, and will be implemented within 30

days of the date of issuance.

Amendment No.: 116

Facility Operating License No. NPF-30: The amendment revised the

Technical Specifications.

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

20857) The August 2, 1996, August 19, 1996, and September 5, 1996,

supplemental letters provided clarifying information and did not change

the original no significant hazards consideration determination. The

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

Safety Evaluationdated October 1, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

Callaway County, Missouri

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

by letters dated July 15, 1996, July 31, 1996, and August 28, 1996.

Brief description of amendment: The amendment would change

Technical Specification (TS) 3/4.3 to support a future modification to

replace existing digital portions of the main steam and feedwater

isolation system (MSFIS) with digital processor equipment and would

authorize revision of the FSAR to include a description of the MSFIS

modification.

Date of issuance: October 1, 1996

Effective date: October 1, 1996, to be implemented prior to startup

from the Callaway Plant Refuel 8.

Amendment No.: 117

Facility Operating License No. NPF-30: The amendment revised the

Technical Specifications and the Final Safety Analysis Report.

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

28619) The July 15, 1996, July 31, 1996 and August 28, 1996

supplemental letters provided additional clarifying information and did

not change the staff's original no significant hazards consideration

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

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

hazards consideration comments received: No.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

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

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: April 4, 1996

Brief description of amendment: The amendment revises the Technical

Specifications regarding the surveillance requirement for control rod

over-travel by moving the specific testing methodology to licensee

administratively controlled documents. Specifically, the amendment

removes the requirement in Specification 4.3.B.1(b) to verify prior to

coupling that the over-travel indicating light is working properly by

withdrawing an uncoupled control rod drive to the over-travel position.

[[Page 55047]]

Date of issuance: September 30, 1996

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

within 30 days.

Amendment No.: 149

Facility Operating License No. DPR-28: Amendment revised the

Technical Specifications.

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

20860) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 30, 1996.No

significant hazards consideration comments received: No

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of application for amendment: August 9, 1996

Brief description of amendment: The amendment changes the

operations manager qualification requirements to allow either of two

alternatives (having held a senior reactor operator's license or having

been certified for equivalent senior reactor operator knowledge) to the

requirement for the operations manager to hold a senior reactor

operator's license.

Date of issuance: October 1, 1996

Effective date: October 1, 1996

Amendment No.: 148

Facility Operating License No. NPF-21: The amendment revised the

Technical Specifications.

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

44350) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: July 3, 1996, as supplemented on

July 23, August 28, and September 16, 1996

Brief description of amendment: The amendment revises Kewaunee

Nuclear Power Plant Technical Specification 4.2.b, ``Steam Generator

Tubes,'' and its associated basis, by revising the acceptance criteria

for indications of tube degradation occurring in the tubesheet crevice

region.

Date of issuance: October 2, 1996

Effective date: October 2, 1996

Amendment No.: 129

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

Technical Specifications.

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

40031) The July 23, August 28, and September 16, 1996, submittals

provided clarifying information that did not change the initial

proposed no significant hazards consideration determination.The

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

Safety Evaluation dated October 2, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

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

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,

Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of application for amendments: May 29, 1996, as supplemented

August 20, 1996

Brief description of amendments: These amendments revise Technical

Specification (TS) Section 15.4.4, ``Containment Tests,'' to

incorporate the provisions of 10 CFR Part 50, Appendix J, ``Primary

Reactor Containment Leakage Testing for Water-Cooled Power Reactors,''

Option B. Revisions have also been made to TS Sections 15.1,

``Definitions,'' 15.3.6, ``Containment System,'' and 15.6,

``Administrative Controls,'' to support the proposed changes to Section

15.4.4.

Date of issuance: October 9, 1996

Effective date: October 9, 1996, to be implemented within 45 days.

Amendment Nos.: Unit 1 - 169 and Unit 2 - 173

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications.

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

34901) The supplemental information did not affect the staff's initial

no significant hazards consideration determination.The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated October 9, 1996.No significant hazards consideration

comments received: No

Local Public Document Room location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

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 for the

amendment 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.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an

[[Page 55048]]

opportunity for public comment. If comments have been requested, it is

so stated. In either event, the State has been consulted by telephone

whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

application for amendment, (2) the amendment to Facility Operating

License, 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 room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By November 22, 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 cro

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