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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice Applications and Amendments to Facility Operating

Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from March 30, 1996, through April 12, 1996. The

last biweekly notice was published on April 10, 1996 (61 FR 15985).

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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 Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, 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 May 24, 1996, the licensee may file a request for a hearing with

respect to issuance of the amendment to the subject facility operating

license and any person whose interest may be affected by this

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

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

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

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

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

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

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

Building, 2120 L Street NW., Washington, DC and at the local public

document room for the particular facility involved. If a request for a

hearing or petition for leave to intervene is filed by the above date,

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

Board Panel, will rule on the request and/or petition; and the

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) The nature of the petitioner's right under the

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

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

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555, Attention: Docketing and Services

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

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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, 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: March 20, 1996.

Description of amendment request: The licensee proposes to relocate

Technical Specification (TS) 3.3.3.2, Movable Incore Detectors, to the

Harris Nuclear Plant Core Operating Limits Report (COLR). Future

changes to the relocated provisions will be evaluated in accordance

with 10 CFR 50.59.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The proposed change will simplify the Technical Specifications,

while implementing the recommendations of the Commission's Final Policy

Statement on TS Improvements. The changes are administrative in nature

and do not involve any modifications to plant equipment or affect plant

operation. Since the TS provisions are being relocated to a licensee-

controlled document, any future changes will be controlled under 10 CFR

50.59. Therefore, there would be no increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The proposed change is a relocation of existing Technical

Specification provisions. It does not involve any physical alterations

to plant equipment or alter the method by which any safety-related

system performs its 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.

The proposed change does not affect any Final Safety Analysis

Report (FSAR) Chapter 15 accident analyses or have any impact on margin

as defined in the Bases to the Technical Specifications. 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: W. D. Johnson, Vice President & Senior

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

North Carolina 27602

NRC Project Director: Eugene V. Imbro

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

Plant, Middlesex County; Northeast Nuclear Energy Company, et al.,

Docket Nos. 50-245, 50-336, 50-423, Millstone Nuclear Power Station,

Units 1, 2, and 3, New London County, Connecticut; and North Atlantic

Energy Service Company, Docket No. 50-443, Seabrook Station, Unit No.

1, Rockingham County, New Hampshire

Date of amendment request: February 1, 1996

Description of amendment request: The amendment request would

revise Section 6 ``Administrative Controls,'' of the Haddam Neck Plant,

Millstone Unit Nos. 1, 2, and 3, and Seabrook Station, Unit 1 Technical

Specifications to reflect several changes in organizational titles. The

proposed changes are administrative title and editorial changes only.

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 (SHC), which is presented below:

* * * The proposed changes do not involve an SHC because the change

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

No design basis accidents are affected by these proposed changes.

The proposed changes are administrative and editorial in nature and are

being proposed to reflect the recently announced organizational changes

which will become effective on February 1, 1996. These changes include:

insertion of the function Chief Nuclear Officer, in lieu of Executive

Vice President--Nuclear; and establishment of a single point of

operational direction for all five units in the position of the Vice

President--Nuclear Operations. This individual is in lieu of the

positions of Vice President--Haddam Neck, Senior Vice President--

Millstone Station, and Executive Director--Nuclear Production. These

latter positions have been eliminated; other changes are: the

appointment of the Haddam Neck Plant Nuclear Unit Director as chairman

of the Haddam Neck PORC [Plant Operations Review Committee]; promotion

of the Shift Supervisor/Shift Superintendent to the position of Shift

Manager; revising the titles of ``additional operator'' and ``auxiliary

operator'' to ``nuclear systems operator''; modifying the phrase

``crewman'' to a gender neutral term ``crewperson'';

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reassignment of the delivery of ISEG [Independent Safety Engineering

Group] reports to the Senior Vice President--Nuclear Safety and

Oversight; and a change to the title of the Seabrook Station Manager to

Station Director. No safety systems are adversely affected by the

proposed changes, and no failure modes are associated with the changes.

Therefore, there is no impact on the probability of occurrence or the

consequences of any accidents previously evaluated.

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

from any accident previously evaluated.

Because there are no changes in the way the plants are operated due

to this administrative change, the potential for an unanalyzed accident

is not created. There is no impact on plant response, and no new

failure modes are introduced. These proposed administrative and

editorial changes have no impact on safety limits or design basis

accidents, and they have no potential to create a new or unanalyzed

event.

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

The changes do not directly affect any protective boundaries nor do

they impact the safety limits for the protective boundaries. These

proposed changes are administrative and editorial in nature. Therefore,

there can be no 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 locations: For the Haddam Neck Plant,

Russell Library, 123 Broad Street, Middletown, CT 06457; for Millstone

Nuclear Power Station, Unit Nos. 1, 2, and 3, Learning Resources

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

Turnpike, Norwich, CT 06360; for Seabrook Station, Unit No. 1, Exeter

Public Library, Founders Park, Exeter, NH 03833.

Attorney for Licensees: 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

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

Station, Unit 1, York County, South Carolina

Date of amendment request: January 26, 1996.

Description of amendment request: The amendment would allow a one-

time change to the Technical Specifications (TS) to allow operation of

the containment purge ventilation system during Modes 3 and 4 during

startup following the forthcoming Unit 1 steam generator replacement

outage. This would alleviate respiratory hazards to personnel who would

enter the containment to perform surveillances during Modes 4 and 3 of

startup operations. Those hazards are expected to result from the

thermal decomposition product gases evolving from the heatup of newly

installed thermal insulation. Operation of the containment purge system

to exhaust these gases would ensure that the air quality meets

applicable standards for personnel safety.

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 activity does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The VP [Containment Purge] System has no interfaces with any

primary system, secondary system, or power transmission system. It has

no interfaces with any reservoir of radioactive gases or liquids. None

of the systems listed above are modified by the activity. In summary,

no ``accident initiator'' is affected with the proposed operation of

the VP System in Mode[s] 3 and 4. For this reason, the activity does

not involve an increase in the probability of an accident previously

evaluated.

Analyses have been performed to determine upper bounds to the

source term, the offsite doses, and the Control Room dose. The results

of that analyses are reported above. Both the source term and the doses

were found to be significantly lower than the results of the

corresponding design basis analyses. No credit was taken for operation

of the annulus ventilation system (VE) in the dose analysis. In

addition, it has been determined that with no credit taken for any heat

transfer from the fuel and cladding to the moderator channels, that

sufficient time would exist for the operators to initiate recovery of

flow from the ECCS [Emergency Core Cooling System] to the reactor core.

The flow required from the ECCS to maintain the core in a coolable

geometry was found to be well within the capacity of any one ECCS pump.

Furthermore, it was determined that convective heat transfer to steam

would be sufficient to prevent release of significant source term or a

significant degree of fuel damage.

For the above reasons, it is determined that operation of the VP

System in Mode 3 or 4 immediately following the steam generator

replacement outage does not involve a significant increase in either

the probability or the consequences of an accident previously

evaluated.

(2) The activity does not create the possibility of a new or

different type of accident from any accident previously evaluated.

As discussed above, no ``accident initiators'' are affected by the

proposed activity. Operation of the VP System proposed for Modes 3 and

4 will be the same as that routinely carried in other modes of

operation. For these reasons, the activity will not create the

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

evaluated.

(3) The activity does not involve a significant reduction in the

margin of safety.

Margin of safety is associated with confidence in the ability of

the fission product barriers (the fuel and fuel cladding, the Reactor

Coolant System pressure boundary, and the containment) to limit the

level of radiation doses to the public. The proposed operation of the

VP System will occur at the end of an extended outage. The level of

decay heat and activity in the reactor is very low compared to the

level of decay heat and activity associated with full power operations.

For this reason, the likelihood of damage to the fuel following a

DBLOCA [design basis loss-of-coolant analysis] occurring during the

proposed purging is reduced, as determined above. Both offsite doses

and doses to the Control Room were found to be small compared to the

limits of 10 CFR [Part] 100 and GDC [General Design Criterion] 19. For

these reasons, the activity 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: 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.

[[Page 18166]]

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

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

Date of amendment request: December 12, 1995.

Description of amendment request: The proposed amendments would

correct an error in the Axial Flux Difference (AFD) Equations to more

accurately reflect the proper AFD limit reduction, which is more

conservative than the literal interpretation of the current Technical

Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

A. The change would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The monitoring of core power distribution and peaking factors is to

ensure accident analysis assumptions such as maximum local pin power at

the initiation of an accident are satisfied, and are not involved in

the initiation or mitigation of any previously evaluated accident.

The proposed change is actually more conservative than the existing

Technical Specification currently being used at McGuire.

B. The change will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

No plant modifications (hardware or control methods) are involved

with this proposed change. The change is simply to correct an error in

the Specification introduced in Amendments 130 (Unit 1) and 112 (Unit

2). The proposed change is more restrictive than the current

specification. No changes are proposed which could create any new

accident scenarios.

C. The proposed change will not involve a significant reduction in

any margin of safety.

The proposed change ensures the margin of safety is properly

maintained by properly reducing (instead of increasing) the Positive

AFD [Axial Flux Difference] limit if a peaking factor exceeds its

surveillance limit. The change is more conservative than the existing

Specification and will ensure the margins of safety are properly

maintained.

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: Atkins Library, University of

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

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. 50-369 and 50-370, McGuire Nuclear

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

Date of amendment request: March 4, 1996.

Description of amendment request: The proposed amendments would

delete the Flow Monitoring System from Technical Specification (TS)

3.4.6.1 and associated surveillance requirements. The TS requires that

either the Containment Floor and Equipment Sump Level System or the

Flow Monitoring System be used to ensure that Reactor Coolant leakage

is maintained within the specified limits. Duke Power does not use the

Flow Monitoring System as a result of documented instrumentation

inaccuracies due to the as-built piping configuration. The existing

piping configuration does not ensure a water solid line which is

necessary for the correct operation of any type of flow

instrumentation. Modification to add a loop seal downstream of the flow

element would be necessary for operability, which would create access

difficulties as well as increase the potential for a radiological

hazard in the form of a CRUD trap.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. This amendment will not significantly increase the probability

or consequence of any accident previously evaluated.

This change will not increase the probability or consequences of an

accident since this Reactor Coolant Leakage Detection instrumentation

is not an accident initiator or mitigator.

This proposed Technical Specification change does not decrease the

number of methods for Reactor Coolant leakage detection. This change

will ensure there are still three distinctly separate methods of

detecting NC [reactor coolant] leakage within the Containment Building.

The first method will be detecting liquid leakage inside Containment

via CFAE [Containment Floor and Equipment] level monitoring. The second

method is detecting an increase in Radiation levels inside Containment

and the third method is detecting steam leakage inside Containment. All

three methods satisfy the diversity requirements listed in Regulatory

Guide 1.45 for detecting a Reactor Coolant leak inside Containment.

The sensitivity requirement listed in Regulatory Guide 1.45 is to

detect a Reactor Coolant leak of one (1) gpm in one (1) hour. The first

method meets this by use of the Sump level monitoring and rate of

increase alarm from this level monitoring device. There are two sumps

inside containment and the levels for both sumps are combined for

detecting a one (1) gpm leak. McGuire uses the Sump Level monitoring to

adequately address liquid leakage detection inside Containment;

therefore, a flow monitoring system on the Sump Discharge line is not

necessary and can be deleted.

The Radiation Monitors are also set up to the required Regulatory

Guide 1.45 sensitivity for detecting Reactor Coolant leakage and are

not designed for SSE [safe-shutdown earthquake] events per the McGuire

FSAR [Final Safety Analysis Report] (see McGuire's Request for

Amendment: Reactor Coolant Leakage Detection Systems, dated March 4,

1996).

The third method for detecting Reactor Coolant leakage is to

monitor Containment Ventilation Condensate Drain Tank (VUCDT) flow, for

which McGuire is also using a level monitor. As in the case of the CFAE

Unit Sump Level monitor, level monitoring for leakage detection is more

reliable than flow monitoring.

2. This amendment will not create the possibility of any new or

different kind of accident not previously evaluated.

The CFAE Flow Monitoring System has no control function, ([i.e.,]

it is only a process monitor). Therefore, its deletion cannot create

the pos[s]ibility of a new or different kind of accident.

3. This amendment will not involve a significant reduction in a

margin of safety.

This proposed Tech Spec change does not decrease the number of

methods for Reactor Coolant leakage detection. This change will ensure

there are still three distinctly separate methods of detecting Reactor

Coolant leakage within the Containment Building.

Tech Spec 3.4.6.1 specifies two Radiation Monitors as two separate

[[Page 18167]]

required methods for Reactor Coolant Leakage Detection with the

Containment Ventilation condensate level monitoring as a backup. The

third method is the Containment Sump level monitoring with the flow

monitoring as a backup.

The new standardized Tech Spec 3.4.15, lists method one as

Containment Sump (Level OR Discharge Flow) Monitoring Device. McGuire

proposes to use a Sump Level monitoring device only. The second method

listed is one Containment Radiation Monitor (either the gaseous or

particulate monitor). McGuire will still have both available. The third

method listed is one Containment air cooler condensate flow rate

monitor for which McGuire plans to also use a level monitor. Liquid,

Radiation, and Steam monitoring will still be accounted for in the Tech

Spec, with the additional requirement of running a Reactor Coolant leak

calculation if any of the methods are inoperable.

Since McGuire is retaining three distinct methods of Reactor

Coolant leakage detection per current TS [technical specification]

requirements (and in agreement with current ISTS [improved standard

technical specification] requirements), the proposed Technical

Specification amendment does not cause any reduction in safety margin.

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: Atkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and

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

Georgia

Date of amendment request: February 21, 1996.

Description of amendment request: The licensee proposes a change to

the Plant Hatch Unit 1 and Unit 2 Technical Specifications. The

proposed revision would change the Drywell Air Temperature Limiting

Condition for Operation (LCO) from less than or equal to 135 deg.F to

less than or equal to 150 deg.F. The proposed change would provide a

margin for the primary containment Drywell Air Temperature LCO when

prolonged summer and high river temperatures are experienced. Also, a

correction to a Final Safety Analysis Report (FSAR) reference would be

made. This typographical error is strictly editorial.

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 probability (frequency of occurrence) of previously evaluated

accidents is not a function of the ambient drywell air temperature.

Instrumentation setpoint calculations were assessed, and the increased

ambient drywell air temperature does not affect any instrumentation

setpoints or allowable values.

The design basis accidents were reevaluated utilizing the increased

drywell air temperature as an initial assumption. The results indicated

that no regulatory limits or equipment design requirements will be

exceeded as the result of the proposed change. Therefore, the change in

drywell air temperature does not result in a significant increase in

the probability or consequences of any previously evaluated accidents.

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

different kind of accident from any accident previously analyzed.

Revising the Drywell Air Temperature LCO does not physically modify the

plant nor does it modify the operation of any existing equipment.

3. The proposed change does not involve a significant reduction in

a margin of safety. Design bases analyses performed utilizing 150 deg.F

as the initial drywell temperature demonstrate that design and

regulatory limits are not exceeded. Equipment in the drywell required

to mitigate the effects of a DBA [design basis accident] is qualified

to operate under environmental conditions expected for an accident.

Analysis results do not affect instrumentation setpoints or

calibration, or accident equipment qualification.

Equipment qualified life is evaluated by an existing program which

uses elevation-dependent drywell temperature rather than bulk average

temperature. Therefore, the margin of safety associated with safety and

other limits identified in the Technical Specifications are not

significantly reduced.

The correction to an FSAR reference is strictly editorial.

Therefore, it meets the three criteria stated above.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513.

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

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

NRC Project Director: Herbert N. Berkow.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: March 28, 1996 (TSCR 234).

Description of amendment request: The proposed amendment modifies

statements in the Technical Specifications and bases to correctly

reflect the reference parameter for anticipatory scram signal bypass.

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. State the basis for the determination that the proposed activity

will or will not increase the probability of occurrence or consequences

of an accident.

This change modifies the terminology in a footnote to a Technical

Specification Table and the bases. The change properly aligns the

footnote and the bases with the FSAR [final safety analysis report] and

the newly revised conservative setpoint which now correctly correlates

the high pressure turbine third stage extraction steam line pressure to

rated reactor thermal power. The change does not modify the function or

operation of the bypass logic. Therefore, the proposed change will not

increase the probability of occurrence or consequences of an accident.

2. State the basis for the determination that the activity does or

does not create the possibility of an accident or malfunction of

equipment of

[[Page 18168]]

a different type than any previously identified in the SAR.

The change does not involve any hardware and does not alter the

functional intent of the pressure switches. The change of the footnote

wording and the bases are primarily administrative and the existing

Technical Specification Limiting Condition for Operation are preserved.

Thus the proposed activity does not create the possibility of an

accident or malfunction of a different type than any previously

identified in the SAR.

3. State the basis for the determination that the margin of safety

as defined in the bases of any Technical Specification is not reduced.

The revised setpoint assures that the anticipatory scram signal

bypass is removed before reaching the Technical Specification limit of

40 percent rated reactor thermal power (during power ascension). Thus,

the margin of safety as stated in the bases of Technical Specification

3.1 is preserved.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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

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

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

NRC Project Director: John F. Stolz.

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

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

Date of amendment request: February 22, 1996 (U-602554)

Description of amendment request: The proposed amendment would

modify Technical Specifications 3.3.8.1, ``Loss of Power

Instrumentation,'' and 3.8.1, ``AC Sources-Operating.'' The proposed

changes would delete the Surveillance Requirement (SR) 3.3.8.1.1 which

requires a channel check for Loss of Power instrumentation and change

Technical Specification Table 3.3.8.1-1 to change the allowable value

for the Degraded Voltage Function (items 1.c and 2.c) from ``[greater

than or equal to] 3762V and [less than or equal to] 3832V'' to

``[greater than or equal to] 3876V.'' The amendment would also change

Technical Specification Table 3.3.8-1 to modify the Division 3 degraded

voltage logic to be the same as Divisions 1 and 2 (i.e., two-out-of-two

rather than three-out-of-three), and increase the steady state voltage

from [greater than or equal to] 3740V to [greater than or equal to]

3870V for SRs 3.8.1.2, 3.8.1.7, 3.8.1.11, 3.8.1.12, 3.8.1.15, 3.8.1.19

and 3.8.1.20.

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) None of the proposed changes involve a significant increase in

the probability or consequences of any accident previously evaluated.

Each of the proposed changes is evaluated against this criteria as

discussed below.

The deletion of the channel check surveillance will result in

discontinuing the recording of information that is not effective in

assessing the capability of the degraded voltage relays to perform

their intended function. Deletion of the channel check does not change

the design or the expected performance of the Loss of Power (LOP)

degraded voltage instrumentation, and therefore, the proposed change

does not impact the intended function of this instrumentation to ensure

adequate voltage for the ECCS equipment during DBA and other non-

accident scenarios. This surveillance provides little added assurance

of relay operability since the relay is normally in a ``non-tripped''

state.

The revision of the Allowable Values for the LOP degraded voltage

and increase in the minimum required voltage for testing diesel

generators will not result in any increase in the probability or

consequences of any accident. The revised Allowable Values will

continue to provide assurance that adequate voltage is available to run

ECCS equipment during DBAs or any other non accident scenarios. With

the emergency bus(es) voltage at or greater than the revised Allowable

Values, the operability of required ECCS equipment is assured. The

revised setpoints for the degraded voltage instrumentation, as

controlled under 10CFR50.59 in the Clinton Power Station Operational

Requirements Manual (ORM), are sufficiently low to assure that the

possibility of spurious trips is minimized.

The planned modification for Division 3 LOP degraded voltage

sensor/relay logic will make Division 3 logic identical to the present

designs for Division 1 and 2. The proposed design for Division 3 will

not result in an increase in the probability of any accident because

the proposed LOP Degraded Voltage logic for Division 3 will be

identical to the proven design of Division 1 and 2. There will not be

an increase in the consequences of an accident because the design of

the LOP Degraded Voltage instrumentation will continue to ensure

adequate voltage for ECCS equipment during any DBA and during non-

accident scenarios.

The proposed footnotes merely assure that the proposed changes

become effective upon installation of the corresponding plant

modifications. Thus, these changes are purely administrative.

Chapter 15 of the Clinton Updated Safety Analysis Report (USAR)

discusses the effects of anticipated process disturbances to determine

their consequences and the capability of the plant to control or

accommodate such events. Subsection 15.2.6 discusses loss of AC power,

including loss of grid voltage. This discussion demonstrates that fuel

design limits and reactor coolant pressure boundary design conditions

are not exceeded. The proposed changes do not affect the discussion nor

the conclusion of this evaluation.

(2) None of the proposed changes create the possibility of a new or

different kind of accident from any accident previously evaluated. Each

of the proposed changes is evaluated against this criterion as

discussed below.

The proposed changes (deletion of the channel check, the revised

Allowable Value for the LOP degraded voltage instrumentation, revision

of the minimum required voltage for the diesel generator (DG)

surveillance, and change of the number of required channels for

Division 3) do not alter the intent or purpose of the degraded voltage

instrumentation. The instrumentation will continue to function to

protect the loads on the emergency bus by switching automatically to

the on site power source when the voltage has been at a degraded

condition for greater than the Allowable Value of the time delay. The

LOP instrumentation provides a responsive actuation (trip) to an

accident or scenario where the protection provided by this function

prevents damage to ECCS equipment during undervoltage (degraded

voltage) conditions on the emergency bus(es). Because the

instrumentation will continue to function to ensure that the emergency

bus voltage for all three divisions is sufficient for the proper

operation of all class 1E equipment down to the 120 volt level, the

proposed change does not create the possibility of a new or different

kind of accident from any accident previously evaluated. The change in

the lower voltage for the DG

[[Page 18169]]

surveillances will not impact the way the surveillances are conducted

because the DGs are run as close to the nominal voltage as possible.

The lower voltage is a criterion for evaluating the surveillance and

the revised lower voltage is adequate for its intended purpose.

(3) None of the proposed changes involve a significant reduction in

a margin of safety. Each of the proposed changes is evaluated against

this criterion as discussed below.

The proposed deletion of the channel check SR 3.3.8.1.1 will not

result in any reduction of the margin of safety because the channel

check is ineffective and the status of the channel will continue to be

apparent to plant personnel because of information provided by other TS

required surveillances. The margin of safety is provided by LOP

instrumentation ensuring the emergency bus(es) have adequate voltage to

support ECCS operability. The proposed revision of the Allowable Value

for the LOP degraded voltage will provide assurance that emergency

bus(es) voltage will be adequate for ECCS loads during DBA and other

non-accident scenarios. These setpoints were determined based on

revised voltage calculations and using an NRC-approved setpoint

methodology. Thus, these changes will not involve any reduction of the

margin of safety. The proposed revision of the number of required

channels for Division 3 will not result in a reduction in a margin of

safety because the proposed Division 3 LOP Degraded Voltage

instrumentation logic will be the same as the proven design of Division

1 and 2. This modification will improve plant maintenance and training

by making Divisions 1, 2 and 3 similar thereby enhancing plant

performance and safety.

Similarly, the proposed revision of the lower voltage limit for

voltage for the DG surveillances (SR 3.8.1.2, SR 3.8.1.7, SR 3.8.1.11,

SR 3.8.1.12, SR 3.8.1.15, SR 3.8.1.19, and SR 3.8.1.20) will assure

that the DGs will be capable of controlling voltage to a range that

will be adequate for the loads on the bus. This value was determined

using revised voltage calculations and is consistent with the proposed

degraded voltage setpoints. None of the proposed changes will 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: Vespasian Warner Public

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

Attorney for licensee: Leah Manning Stetener, Vice President,

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

Decatur, Illinois 62525.

NRC Project Director: Gail H. Marcus.

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

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

Date of amendment request: February 22, 1996 (U-602551).

Description of amendment request: The proposed amendment would

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

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

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

apart from the core beltline region of the reactor vessel.

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 results in a specific pressure and

temperature (P/T) limit curve for the bottom head during vessel

pressure testing evolutions, while the P/T limits for the remaining

balance of reactor pressure vessel regions are unchanged. The limits

for the bottom head region, which are only applicable during vessel

system pressure or leak testing, were developed consistent with

Regulatory Guide 1.99, Revision 2; 10CFR50, Appendix G; ASME Section

III, Appendix G; and Welding Research Council (WRC) Bulletin 175.

Additionally, the proposed change does not result in a change to the

way in which the hydrostatic pressure tests are performed. That is,

conformance to the P/T limits specified in Technical Specification

Figure 3.4.11-1 with the proposed bottom head P/T limits incorporated,

will continue to provide protection against brittle fracture of the

vessel system during required testing so that vessel integrity is

maintained. Therefore, this proposed change does not result in an

increase in the probability or consequences of any accident previously

evaluated.

(2) The proposed change does not result in any change to the plant

or the way in which the hydrostatic pressure tests are performed. As a

result, no new failure modes are introduced. Therefore, the proposed

change cannot create the possibility of a new or different kind of

accident from any accident previously evaluated.

(3) The new P/T limit curve for the bottom head has been developed

consistent with Regulatory Guide 1.99, Revision 2; 10CFR50, Appendix G;

ASME Section III, Appendix G; and Welding Research Council (WRC)

Bulletin 175. All other regions of the reactor pressure vessel retain

their applicability to appropriate and previously approved P/T limit

curves which are based on the same methodology. Conformance to the P/T

limit curves, with the proposed changes incorporated, will continue to

provide adequate margins of safety against brittle fracture of the

reactor vessel. Therefore, this proposed change does not result in a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

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

Attorney for licensee: Leah Manning Stetener, Vice President,

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

Decatur, Illinois 62525.

NRC Project Director: Gail H. Marcus.

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

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

Date of amendment request: February 22, 1996 (U-602522)

Description of amendment request: The proposed amendment would

change Technical Specification 3.3.4.1, ``End of Cycle Recirculation

Pump Trip (EOC-RPT) Instrumentation,'' by deleting Surveillance

Requirement (SR) 3.3.4.1.6. The SR requires the reactor recirculation

pump trip breaker interruption time to be determined at least once per

60 months.

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) End of cycle recirculation pump trip (EOC-RPT) actuation in

response to main generator load rejection and main turbine trip events

has previously been evaluated in Chapter 15 of Clinton Power Station

(CPS) Updated Final

[[Page 18170]]

Safety Analysis. The proposed change does not affect the initiators of

any of these events. In addition, the possibility of failure of the

EOC-RPT breaker to mitigate these events has not been increased because

there has been no change in design and no change to the plant. Deleting

the requirement to periodically measure the breaker arc suppression

time will not impact the EOC-RPT breakers' capability of performing

their intended function because CPS will continue to perform

inspections, testing and maintenance that supports breaker operation as

intended and provides assurance that breaker interruption time will be

within limits. Thus, the EOC-RPT breaker trip may be expected to

operate as before to mitigate pressurization transient effects.

The EOC-RPT breaker trip is also assumed to occur in the analyses

for the loss of feedwater heating, feedwater controller failure,

pressure regulator failure, recirculation flow control failure, and

recirculation pump seizure events. However, the EOC-RPT breaker trip is

not an initiator or mitigating feature for these events. The proposed

change cannot therefore impact the probability or consequences for

these events. Nonetheless, the EOC-RPT breaker trip may be assumed to

function as before for these scenarios.

For scenarios where the EOC-RPT breaker trip could initiate an

event (i.e., inadvertent recirculation pump trip events), the

probability of occurrence is not increased. The design and operation of

the EOC-RPT system has not been changed, and therefore, the

consequences resulting from the EOC-RPT breaker trip are unchanged.

Based on the above, neither the probability nor the consequences of

any accident previously evaluated have been increased.

(2) As noted above, the EOC-RPT breakers will continue to function

as before. The proposed change involves no design change or physical

change in the plant. Therefore, previous accident analyses are

unchanged. Further, no new operations or testing is involved. On this

basis, no new failure modes are introduced. Therefore, this proposed

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

accident from any accident previously evaluated.

(3) This proposed change does not involve a significant reduction

in a margin of safety. The capability of the EOC-RPT breaker trip to

provide additional insertion of negative reactivity for mitigating

design-basis events remains unchanged. That is, the EOC-RPT will

continue to be capable of reducing the peak reactor pressure and power

resulting from turbine trip or generator load rejection transients,

thus providing additional margin to core thermal MCPR Safety Limits.

The margin of safety is assured by the EOC-RPT breaker trip

occurring within established limits such that the overall system

performs its intended safety function within the time analyzed for the

system safety response. No system time limit change is proposed. The

robust design of the breakers, combined with continued performance of

vendor-recommended testing and maintenance that ensures proper

mechanical and electrical performance of the breakers, will continue to

provide assurance that breaker interruption time is within the

acceptable limit. Therefore, there is 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: Vespasian Warner Public

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

Attorney for licensee: Leah Manning Stetener, Vice President,

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

Decatur, Illinois 62525.

NRC Project Director: Gail H. Marcus.

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

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

Date of amendment request: February 22, 1996 (U-602549).

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.6.5.1, ``Drywell,'' to allow

drywell bypass leakage tests to be performed at intervals of up to ten

years based, in part, on the demonstrated performance of the drywell

barrier with respect to leak tightness. The proposed amendment would

also revise TS 3.6.5.2, ``Drywell Air Lock,'' to extend the testing

intervals for the surveillances on drywell air lock overall leakage and

interlock operability, relocate the specific leakage limits on the air

lock barrel and door seals to the TS Bases, relocate the requirement to

pressurize the drywell air lock to 19.7 psid prior to performance of

the overall drywell air lock leakage test to the TS Bases, and other

administrative changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

(1) The proposed changes do not involve a change to the plant

design or operation. As a result, the proposed changes do not affect

any of the parameters or conditions that contribute to initiation of

any accidents previously evaluated. Therefore, the proposed changes

cannot increase the probability of any accidents previously evaluated.

The proposed changes do potentially affect the leaktight integrity

of the drywell, a structure used to mitigate the consequences of a loss

of coolant accident (LOCA). The function of the drywell is to force the

steam released from a LOCA through the suppression pool, limiting the

amount of steam released to the primary containment atmosphere. This

serves to limit the containment pressurization due to the LOCA. The

leakage of the drywell is limited to ensure that the primary

containment does not exceed its design limits of 185 deg.F and 15 psig.

Because the proposed change to replace the current 18-month frequency

for performing drywell bypass leakage tests (DBLRTs) with a

performance-based frequency does not alter the plant design, the

proposed change does not directly result in an increase in the drywell

leakage. However, decreasing the test frequency can increase the

probability that a large increase in drywell bypass leakage could go

undetected for an extended period of time. This potential has been

evaluated, and Illinois Power has determined that the proposed change

to the DBLRT frequency will not result in the potential for undetected,

large increases in leakage, as further discussed below.

There are several potential drywell bypass leakage paths. These

include potential cracks in drywell concrete structure, the drywell

vacuum breakers, and various penetrations through the drywell

structure. Based on the results of the structural integrity test

conducted at the design pressure of 30 psig as part of the

preoperational test program, additional cracking of the drywell is not

expected during the remaining life of the plant. Ventilation and piping

penetrations (including the drywell vacuum breaker penetrations) are

designed to ASME Code Class 2 and Seismic Category 1 requirements.

These penetrations are typically designed with two isolation valves in

series with one valve in the drywell and another either outside primary

containment or in the wetwell. Technical Specification (TS)

Surveillance Requirements (SRs) require, as applicable, periodic

verification of drywell isolation valve

[[Page 18171]]

position, stroke time, and automatic isolation capability. High energy

lines that extend into the wetwell, such as the main steam lines and

feedwater lines, are encapsulated by guard pipes to direct energy back

into the drywell in case of a piping rupture. Electrical penetrations

are sealed with a high strength/density material that will prevent

leakage, as well as provide radiation shielding.

The proposed changes for the drywell air lock involve relocation of

the separate limits on the drywell air lock barrel and seal leakage

rates to the TS Bases, relocation of the requirement to pressurize the

air lock to 19.7 psid prior to performance of the air lock overall

(barrel) leakage test, and changing the frequency for these tests from

18 months to 24 months. While the proposed changes will eliminate

separate TS limits on leakage of the drywell air lock, the overall

drywell bypass leakage TS limit (which includes leakage through the air

lock) is not affected by this proposed change. The limiting scenario

for drywell bypass leakage is a small break LOCA which results in

drywell pressures of approximately 3 psid. Only a large break LOCA can

create drywell pressures of 19.7 psid. For this event, the allowable

drywell bypass leakage rate is over eight times larger than for a small

break LOCA. Thus, relocation of these requirements to the TS Bases will

continue to provide adequate control of these requirements. The

proposed air lock overall leakage rate testing frequency is consistent

with the guidance for testing primary containment air locks in Nuclear

Energy Institute (NEI) 94-01, ``Industry Guideline for Implementing

Performance-Based Option of 10CFR50, Appendix J.'' The drywell air lock

is tested in a manner similar to the primary containment air locks,

even though the drywell air lock is not a direct leakage path from

primary containment and, therefore, 10CFR50, Appendix J test

requirements do not apply. The drywell air lock's use is limited during

plant operation due to radiation and temperature in the drywell. Since

sufficient confidence in the door's sealing capability is assured

considering past performance and the air lock door usage is very low

throughout an operating cycle, it is justified to allow performance of

these tests at refueling-outage intervals, whether the unit is on a 18-

month or a 24-month refueling cycle.

Operational experience has shown that the leak tightness of the

drywell has been maintained well below the allowable leakage limits at

Clinton Power Station. The TS limit of 10% of the design [maximum

allowable leakage path area] provides a large margin for degradation.

Drywell performance to date suggests that drywell degradation, even

with a ten-year interval between tests, will not exceed this margin.

The most recent DBLRT performed during the fourth refueling outage (RF-

4) measured a drywell bypass leakage rate of 0.07% of the design limit.

An analysis was also conducted to determine the potential risk to

the public from unacceptable drywell bypass leakage going undetected as

a result of the proposed change. Based on this probabilistic risk

analysis, for several different accident scenarios, the risk of

radioactivity release from containment was found to be insignificant.

Based on the above, Illinois Power has concluded that the proposed

changes will not result in a significant increase in the consequences

of any accident previously evaluated.

(2) The proposed change does not involve a change to the plant

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

any of the parameters or conditions that could contribute to initiation

of any accidents. Drywell bypass leakage cannot, of itself, create an

accident. Thus, it has been concluded that the proposed change cannot

create the possibility of an accident not previously evaluated.

(3) The NRC has provided standards for determining whether a no

significant hazards consideration exists as stated in 10CFR50.92(c).

These proposed changes involve the withdrawal of operating restrictions

previously imposed because acceptable operation of the Mark III primary

containment design had not been demonstrated at the time of initial

licensing. As published in the Federal Register (FR) regarding no

significant hazards consideration criteria, granting of a relief based

upon demonstration of acceptable operation from an operating

restriction that was imposed because acceptable operation had not yet

been demonstrated does not involve a significant hazards consideration

(reference 48 FR 14870).

The proposed change only affects the frequency of measuring the

drywell bypass leakage rate and does not change the bypass leakage rate

limit. The proposed change could potentially increase the probability

that a large increase in drywell bypass leakage could go undetected for

an extended period of time. However, operational experience has shown

that the leaktightness of the drywell has been maintained well below

the allowable leakage limits. In addition, there are TS surveillances

which require, as applicable, periodic verification of drywell

isolation valve position, stroke time, and automatic isolation

capability. Further, qualitative methods (such as periodic verification

that the drywell pressurizes, which ensures that the drywell leak rate

is less than the instrument air leak and usage rates) are available to

provide assurance that the drywell leakage rate is being maintained

within limits. The Clinton Power Station TS require the drywell leakage

rate measured during DBLRTs to be less than or equal to 10% of the

design limit. This request does not affect this required margin. Nor

does it affect the existing margin between the primary containment

design pressure and the actual pressure at which primary containment

would fail.

With respect to proposed changes to the drywell air lock overall

leakage testing and interlock testing requirements, the proposed leak

test frequencies are consistent with the guidance for testing primary

containment air locks in NEI 94-01. Due to the limited use of the

drywell air locks during plant operation, it is justified to allow

performance of interlock operability testing on a refueling outage

basis, whether the unit is on an 18-month or a 24-month refueling

cycle. The separate limits on the drywell air lock and barrel are being

relocated from the TS, these limits are being controlled under

10CFR50.59 and the TS Bases Control program of TS 5.5.11. Leakage

through these pathways will continue to be a part of the overall

drywell bypass leakage limited by LCO 3.6.5.1.

An analysis was also conducted to determine the potential risk to

the public from the proposed change. Based on this probabilistic risk

analysis, for several different accident scenarios, the risk of

radioactivity release from containment was found to be insignificant.

As a result, Illinois Power has concluded that the proposed changes

will continue to assure that the drywell bypass leakage will be within

design limits if challenged and therefore, will not result in a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

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

[[Page 18172]]

Attorney for licensee: Leah Manning Stetener, Vice President,

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

Decatur, Illinois 62525.

NRC Project Director: Gail H. Marcus.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

Date of amendment requests: February 26, 1996 (AEP:NRC:1071U).

Description of amendment requests: The proposed amendments would

modify the technical specifications (TS) to increase the current limit

on nominal fuel assembly enrichment for new, Westinghouse-fabricated,

fuel stored in the new fuel storage racks from 4.55 weight percent

uranium-235 isotope to 4.95 weight percent uranium-235 isotope with

certain provisions. Also, TS 5.6.2 would be reformatted similar to that

used in the standard TS (NUREG-1431, Rev. 1).

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:

Per 10 CFR 50.92, a proposed amendment will not involve a

significant hazards consideration if the proposed amendment does not:

(1) involve a significant increase in the probability or

consequences of an accident previously evaluated,

(2) create the possibility of a new or different kind of accident

from any accident previously evaluated, or

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

Criterion 1

The proposed changes will not involve a significant increase in the

probability of an accident previously evaluated because similar

administrative controls to those presently used to identify new fuel

storage rack inventory and compliance with T/S limits will be used.

There are no physical changes to the plant associated with this T/S

change. The consequences of an accident previously evaluated will not

be increased because the reactivity of the fuel stored in the new fuel

storage racks under the proposed T/S limits will be no greater than the

reactivity of fuel stored in the new fuel storage racks presently

allowable under the current T/S limits.

Criterion 2

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

different kind of accident from any accident previously evaluated

because the changes will involve no physical changes to the plant nor

any changes in plant operations. Furthermore, the reactivity of the

fuel stored in the new fuel storage racks under the proposed T/S limits

will be no greater than the reactivity of fuel stored in the new fuel

storage rack presently allowable under the current T/S limits.

Criterion 3

The proposed amendment will not involve a significant reduction in

a margin of safety because the reactivity of the fuel stored in the new

fuel storage racks under the proposed T/S limits will be no greater

than the reactivity of fuel stored in the new fuel storage racks

presently allowable under the current T/S limits.

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. In addition, the reformatting of TS 5.6.2 is a purely

administratiave change having no effect on the physical plant or its

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Mark Reinhart, Acting.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

Date of amendment requests: February 29, 1996 (AEP:NRC:1232).

Description of amendment requests: The proposed amendments would

revise the technical specifications to reduce the boric acid

concentration in the boric acid storage system from approximately 12

percent to approximately 4 percent by weight. Related changes are also

proposed to increase the minimum required flow rate in action

statements for certain affected TS and add an additional surveillance

requirement for this flow rate, and decrease the minimum temperature

requirement in certain affected TS to 63 deg.F. The bases section is

also updated to reflect these proposed changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Per 10 CFR 50.92, a proposed change does not involve a significant

hazards consideration if the change does not:

1. involve a significant increase in the probability or

consequences of an accident previously evaluated,

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

from any accident previously evaluated, and

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

Criterion 1

Does the change involve a significant increase in the probability

or consequences of an accident previously evaluated?

NO. The BAST [boric acid storage tank] water volume and boron

concentration were not credited in any Chapter 14 safety analysis.

Therefore, no change in the probabilities of the accident analysis will

result from the BAST water volume and boron concentration change. In

addition, since the BAST water volume and boron concentration are not

taken into consideration in any safety analysis, the consequences of an

accident previously evaluated in the FSAR [final safety analysis

report] are not increased. The heat tracing system is currently only

necessary to prevent precipitation of existing high boric acid

concentration in the plant systems. The reduction in boron

concentration in this proposal eliminates the need for the heat tracing

system. The existence of the heat tracing system was not part of any

safety analysis and disabling of the heat tracing system will not

result in a significant increase in the probability or consequences of

an accident previously evaluated.

Criterion 2

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

of accident from any accident previously evaluated?

NO. Since the minimum required water flow from the boric acid

storage system to the reactor coolant system was increased to

counteract any possible operational transients, as shown in Attachment

4 [of the application], the change in BAST water volume and boron

concentration and disabling of the heat tracing system do not create

the possibility of an accident which is different from any already

evaluated in the FSAR. No new or different failure modes have been

defined for any system or component nor has any new limiting single

failure been identified.

[[Page 18173]]

Criterion 3

Does the change involve a significant reduction in a margin of

safety?

NO. The margin of safety requirements are not affected by the

removal of the heat tracing system and the reduction of the boric acid

concentration in the boric acid storage system. The required flow paths

and borated water sources are unaffected by this proposal. The required

quantity of borated water is still available based upon the performed

evaluation, and appropriate surveillance requirements ensure the

ability to deliver this borated water. The reduction of the boric acid

concentration in the BASTs will ensure that the boric acid remains in

solution at the normal room temperature in the auxiliary building. With

the above changes, there will be a net improvement in system

reliability and accordingly the proposed changes do not affect 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 requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Mark Reinhart, Acting.

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

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

California

Date of amendment requests: March 13, 1996.

Description of amendment requests: The proposed amendments would

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

Power Plant, Unit Nos. 1 and 2 to revise TS 4.0.5, ``Surveillance

Requirements,'' to delete reference to prior NRC approval for written

relief from the Inservice Inspection (ISI) and Inservice Testing

Program (IST) requirements and to add ASME Section XI definition of

``Biennially or every 2 years--At least once per 731 days'' in TS

4.0.5b.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes implement the NRC's recommendation contained

in NUREG-1482, ``Guidelines for Inservice Testing Programs at Nuclear

Power Plants,'' endorsed by Generic Letter

89-04, Supplement 1, ``Guidance on Developing Acceptable Inservice

Testing Programs.'' The changes are consistent with 10 CFR 50.55a,

``Codes and Standards,'' which does not prohibit the implementation of

relief from ASME Section XI requirements prior to specific written

approval when those changes are found acceptable by change process

specified in 10 CFR 50.59, ``Changes, Tests and Experiments.'' The

proposed changes are administrative in nature and do not involve any

modifications to any plant equipment or affect plant operation.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

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

different kind of accident from any accident previously evaluated.

The proposed changes are administrative in nature, do not involve

any physical alterations to any plant equipment, and cause no change in

the method by which any safety-related system performs its 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 changes do not involve a significant reduction in a

margin of safety.

The proposed changes do not alter the basic regulatory requirements

and do not affect any safety analyses.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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

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

NRC Project Director: William H. Bateman.

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

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

California

Date of amendment request: April 3, 1996.

Description of amendment request: The proposed amendments would

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

Power Plant, Unit Nos. 1 and 2 to revise Technical Specifications 3/

4.7.5, ``Control Room Ventilation System,'' 3/4.7.6, ``Auxiliary

Building Safeguards Air Filtration System,'' and 3/4.9.12, ``Fuel

Handling Building Ventilation System,'' to clarify the testing

methodology utilized by PG&E to determine the operability of the

charcoal and high-efficiency particulate air (HEPA) filters in the

engineering safeguards features (ESF) air handling units.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The charcoal testing protocol changes will not affect system

operation or performance, nor do they affect the probability of any

event initiators. These changes do not affect any engineered safety

features actuation setpoints or accident mitigation capabilities. The

new charcoal adsorber sample laboratory testing protocol more

accurately demonstrates the required performance of the adsorbers in

the control room ventilation system and auxiliary building safeguards

air filtration system following a design basis loss of coolant accident

or in the fuel handling building ventilation system following a fuel

handling accident outside containment. The decontamination efficiencies

used in the offsite and control room dose analyses are not affected by

these changes. Therefore, offsite and control room dose analyses are

not affected by this change, and all offsite and control room doses

will remain within the limits of 10 CFR 100 and 10 CFR 50, Appendix A,

General Design Criterion (GDC) 19.

[[Page 18174]]

The requirements of ANSI N510-1980 encompass the requirements of

ANSI N510-1975, which is referenced in Regulatory Guide (RG) 1.52, as

it applies to testing at Diablo Canyon Power Plant (DCPP).

Consequently, revising the Technical Specifications (TS) to reference

ANSI N510-1980 will have no effect on filter testing.

The proposed changes are consistent with the new Standard Technical

Specifications (NUREG-1431, Rev. 1).

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

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

different kind of accident from any accident previously evaluated.

The changes to the charcoal sample testing protocol will not affect

the method of operation of the system. The proposed changes only affect

the testing parameters for the charcoal samples. No new or different

accident scenarios, transient precursors, failure mechanisms, or

limiting single failures will be introduced as a result of these

changes.

The requirements of ANSI N510-1980 encompass the requirements of

ANSI N510-1975, which is referenced in RG 1.52, as it applies to

testing at DCPP. Consequently, revising the TSs to reference ANSI N510-

1980 will have no effect on filter testing.

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

new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction in

a margin of safety.

The changes in charcoal sample testing protocol will not affect

system performance or operation. The decontamination efficiencies used

in the offsite and control room dose analyses are not affected by these

changes. Therefore, offsite and control room dose analyses are not

affected by this change, and all offsite and control room doses will

remain within the limits of 10 CFR 100 and 10 CFR 50, Appendix A, GDC

19.

The requirements of ANSI N510-1980 encompass the requirements of

ANSI N510-1975, which is referenced in RG 1.52, as it applies to

testing at DCPP. Consequently, revising the TSs to reference ANSI N510-

1980 will have no effect on filter testing.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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

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

NRC Project Director: William H. Bateman.

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

Plant, Unit 3, Humboldt County, California

Date of amendment request: March 13, 1996.

Description of amendment request: The proposed amendment would

revise the Humboldt Bay Power Plant (HBPP), Unit 3, Technical

Specifications (TS) by incorporating position changes to reflect a

proposed plant staff reorganization. The TS changes proposed are as

follows:

(1) TS Section VII.C.2.c and VII.D.1.b--change the position title

from ``Power Plant Engineer'' to ``Senior Power Production Engineer.''

(2) TS Section VII.C.2.d--change the position title from ``Senior

Chemical and Radiological Engineer'' to ``Senior Radiation Protection

Engineer.''

(3) TS Section VII.C.2.e and VII.D.1.b--change the position title

from ``Maintenance Planner'' to ``Supervisor of Maintenance.''

(4) TS Section VII.C.2.g and VII.D.1.b--add the position of

``Assistant Plant Manager/Power Plant Engineer.''

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 administrative and organizational changes provide

editorial corrections and reflect the proposed HBPP and current NRC

organizations. These changes do not affect the operating methodology of

HBPP, and they are not related to the probability or consequences of an

accident previously evaluated.

Therefore, the proposed changes do 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 revisions to the HBPP TS are organizational and

administrative in nature, and do not change the method by which any

safety-related system performs its function.

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

new of different kind of accident from any accident previously

evaluated.

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

safety?

The proposed changes have no effect on the current operating

methodologies or actions that govern plant performance. In addition,

the proposed changes do not affect the margin of safety associated with

parameters for any accident analysis.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the analysis of the licensee and, based

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Humboldt County Library, 1313

3rd Street, Eureka, California 95501.

Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas

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

NRC Project Director: Seymour H. Weiss.

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

Callaway County, Missouri

Date of application request: February 9, 1996, as superseded by

letter dated March 22, 1996.

Description of amendment request: The amendment would revise

Technical Specification (TS) Definition 1.7, TS 3/4.6, TS 6.8, and

their associated bases to directly reference Regulatory Guide 1.163 as

required by 10 CFR 50, Appendix J, Option B, for the Type A containment

integrated leak rate tests (ILRTs) and the Type B and C local leak rate

tests (LLRTs).

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

[[Page 18175]]

consideration, which is presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes to TS 1.7e, 4.6.1.1, 3/4.6.1.3, Bases 3/

4.6.1.1 and the program addition to TS 6.8.4g have no effect on plant

operation. The proposed changes only provide mechanisms within TS for

implementing a performance-based methodology for determining the

frequency of leak rate testing, as allowed by the NRC. The test type,

method, and acceptance criteria will not be changed. Containment

leakage will continue to be maintained within the required limits.

Based on industry and NRC evaluations performed in support of

developing Option B, these changes potentially result in a minor

increase in the consequences of an accident previously evaluated due to

the increased testing intervals. However, the proposed changes do not

result in an increase in the core damage frequency since the

containment system is used for mitigation purposes only.

Directly referencing the Containment Leakage Rate Testing Program

for Containment ILRT and LLRT requirements does not involve any

modification to plant equipment or affect the operation or design basis

of the containment. Leakage rate testing is not a precursor to or an

initiating event for any accident.

Therefore, these changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

The proposed changes only allow for implementation of 10 CFR 50,

Appendix J, Option B and do not involve any modifications to any plant

equipment or affect the operation or design basis of the containment.

The proposed changes do not affect the response of the containment

during a design basis accident.

3. The proposed change does not involve a significant reduction in

a margin of safety.

The proposed changes do not affect or change a safety limit, any

limiting condition for operation or affect plant operations. The

changes only implement the Appendix J, Option B test frequencies that

have been determined by NRC not to involve a safety concern. The

testing methods, acceptance criteria and bases are not changed and

still provide assurance that the containment will provide its intended

function.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: March 21, 1996.

Description of amendment request: The proposed changes to the

Technical Specifications (TS) for the North Anna Power Station, Units

1&2 (NA-1&2) would clarify the requirements for testing charcoal

adsorbent in the Waste Gas Charcoal Filter System, the Control Room

Emergency Habitability System, and the Safeguards Area Ventilation

System. No change in the testing is being proposed, only clarification

of the description of the required testing in TS 3/4.6.4.3, 3/4.7.7.1,

and 3/4.7.8.1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed Technical Specifications changes will revise

Surveillance Requirements for the charcoal adsorbent in the Waste Gas

Charcoal Filter System (TS 3/4.6.3.), Control Room Emergency

Habitability System (TS 3/4.7.7.2), and the Safeguards Area Ventilation

System (TS 3/4.7.8.1) to reflect the current testing methodology for

new and used carbon adsorbent. These proposed changes specify ASTM D

3803-1979 as the laboratory testing standard for both new and used

charcoal adsorbent for the ventilation system identified above.

Virginia Electric and Power has evaluated the proposed Technical

Specification changes to the North Anna Units 1 and 2 Technical

Specifications against the Significant Hazards Criteria of 10 CFR 50.92

and determined that the changes do not involve any significant hazard

for the following reasons:

1. The probability or consequences of an accident previously

evaluated is not significantly increased.

The proposed changes are administrative in nature in that the

changes only explicitly specify the current testing methodology for

charcoal adsorbent. The proposed changes will not affect system

operation or performance, nor do they affect the probability of any

event initiators. These changes do not affect any Engineered Safety

Features actuation setpoints or accident mitigation capabilities.

Therefore, the proposed changes will not significantly increase the

consequences of an accident or malfunction of equipment important to

safety previously evaluated in the UFSAR.

2. The possibility of an accident or a malfunction of a different

type than any previously evaluated is not created.

The proposed changes only clarify the requirements for charcoal

testing and will not affect the method of operation of the ventilation

systems. Furthermore, the proposed changes are only intended to clarify

the existing requirements to explicitly specify the current test

methodology. No new or different accident scenarios, transient

precursors, failure mechanisms, or limiting single failures will be

introduced as a result of these changes. Therefore, the possibility of

a new or different kind of accident other than those already evaluated

will not be created by this change.

3. The margin of safety has not been significantly reduced.

The proposed changes which represent the current laboratory testing

methodology for charcoal adsorber samples, demonstrates the required

performance of the adsorbent following a design basis LOCA or Fuel

Handling Accident. Changing the Technical Specification to clarify the

methodology for charcoal sample testing will not affect system

performance or operation.

Therefore, these changes will not result in a significant reduction

in any margin of safety.

Based on the above discussions, it has been determined that the

requested Technical Specification changes do not involve a significant

increase in the probability or consequences of an accident or other

adverse condition over previous evaluations; or create the possibility

of a new or different kind of accident or condition over previous

evaluation; or involve a significant reduction in a margin of safety.

Therefore, the requested license amendment does not involve a

significant hazards consideration.

[[Page 18176]]

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: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219

NRC Project Director: Eugene V. Imbro.

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.

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

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

Date of amendment request: March 21, 1996.

Brief description of amendments: The amendments provide changes to

Technical Specifications (TS) for CR3 relating to the Once Through

Steam Generator's (OTSG's) tube inspection acceptance criteria, and

repair limit for removing steam generator tubes from service. The

proposed TS change would be applicable for one cycle duration, and only

to Inter-Granular-Attack (IGA) degradation mechanism in a limited

region of the OTSG.

Date of publication of individual notice in Federal Register: March

28, 1996 (61 FR 13888)

Expiration date of individual notice: April 29, 1996.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629.

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

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

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

County, Texas

Date of amendment request: May 30, 1995, as supplemented by letter

dated February 8, 1996.

Description of amendment request: The proposed amendment would

increase the spent fuel pool heat load licensing basis to provide

greater flexibility for normal refueling practices.

Date of individual notice in the Federal Register: April 3, 1996

(61 FR 14832)

Expiration date of individual notice: May 3, 1996.

Local Public Document Room location: Wharton County Junior College,

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

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,

and 3, Maricopa County, Arizona

Date of application for amendments: December 20, 1995.

Brief description of amendments: These amendments change the

instrument setpoint for the reactor trip and main steam isolation

signal actuation on low steam generator pressure from greater than or

equal to 919 psia with an allowable value of 911 psia to 895 psia with

an allowable value of greater than or equal to 890 psia.

Date of issuance: April 5, 1996.

Effective date: April 5, 1996, to be implemented within 45 days of

issuance.

Amendment Nos.: Unit 1-105; Unit 2-97; Unit 3-77.

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

amendments revised the Technical Specifications.

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

FR 7544) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 5, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

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

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

Maryland

Date of application for amendments: November 1, 1995 as

supplemented on December 1, 1995.

Brief description of amendments: The amendments reflect the new

plant electrical distribution configuration, surveillance and limiting

condition for operation of the new safety-related (SR) emergency diesel

generator (EDG), the increased electrical capacities for the two of the

three existing SR EDGs, the increased EDG fuel oil storage capacity,

and the fire protection system for the

[[Page 18177]]

new EDG building. The remaining existing SR EDG will be upgraded during

the Unit No. 2 refueling outage scheduled for the spring of 1997.

Date of issuance: April 2, 1996.

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

30 days.

Amendment Nos.: 214 and 191.

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

revised the Technical Specifications.

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

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

contained in a Safety Evaluation dated April 2, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

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

and 2, Will County, Illinois

Date of application for amendments: December 6, 1995, as

supplemented February 27, 1996, and March 28, 1996.

Brief description of amendments: The amendments modify the

technical specifications to replace the existing scheduling

requirements for overall integrated and local containment leakage rate

testing with a requirement to perform the testing in accordance with 10

CFR Part 50, Appendix J, Option B. Option B allows test scheduling to

be adjusted based on past performance.

Date of issuance: April 4, 1996.

Effective date: April 4, 1996.

Amendment Nos.: 81, 81, 73, and 73.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

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

FR 7547) The February 27, 1996, and March 28, 1996, supplements

modified the Technical Specification pages to be more consistent with

the published guidance, were within this scope of the initial notice,

and did not affect the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated April 4, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: For Byron, the Byron Public

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

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

Wilmington, Illinois 60481.

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

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

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

and 2, Will County, Illinois

Date of application for amendments: October 3, 1995, as

supplemented on February 21, 1996, and April 2, 1996.

Brief description of amendments: The amendments revise the

Technical Specifications (TS) to implement ten of the line-item TS

improvements recommended in Generic Letter (GL) 93-05, ``Line-Item

Technical Specifications Improvements to Reduce Surveillance

Requirements for Testing During Power Operation,'' dated September 27,

1993. The amendments also include editorial changes on the affected TS

pages.

Date of issuance: April 10, 1996.

Effective date: April 10, 1996.

Amendment Nos.: 82, 82 and 74, 74.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

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

FR 58397). The February 21, 1996, and April 2, 1996, submittals did not

change the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 10, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: For Byron, the Byron Public

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

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

Wilmington, Illinois 60481.

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

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

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

and 2, Will County, Illinois

Date of application for amendments: May 17, 1995, as supplemented

by letters dated January 17, March 8, March 18, April 4 and April 9,

1996.

Brief description of amendments: The amendments revised the

Facility Operating Licenses and the technical specifications to permit

the steam generator tubes to be repaired using the tungsten inert gas

welded sleeve process developed by ABB-Combustion Engineering and

remove references to the kinetically welded sleeving process.

Date of issuance: April 12, 1996.

Effective date: April 12, 1996.

Amendment Nos.: 83, 83, 75, and 75.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised licenses and the Technical Specifications.

Date of initial notice in Federal Register: July 5, 1995 (60 FR

35064) The additional submittals provided information that did not

change the initial proposed no significant hazards consideration

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

contained in a Safety Evaluation dated April 12, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: For Byron, the Byron Public

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

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

Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units

1 and 2, Rock Island County, Illinois

Date of application for amendments: September 1, 1995, for Dresden

and September 20, 1995, for Quad Cities.

Brief description of amendments: This application upgrades the

current custom Technical Specifications (TS) for Dresden and Quad

Cities to the Standard Technical Specifications contained in NUREG-

0123, ``Standard Technical Specification General Electric Plants BWR/

4.'' This application upgrades only Section 6.0, ``Administrative

Controls.''

Date of issuance: April 2, 1996.

Effective date: Immediately, to be implemented no later than June

30, 1996.

Amendment Nos.: 149, 143, 170, and 166.

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30:

The amendments revised the Technical Specifications.

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

FR 48728) for Dresden and October 5, 1995 (60 FR 52226) for Quad

Cities. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 2, 1996.

[[Page 18178]]

No significant hazards consideration comments received: No.

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021.

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

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

Date of application for amendments: January 18, 1996, as

supplemented on March 1, March 22, March 26, and April 3, 1996.

Brief description of amendments: The amendments change the

setpoints for the automatic primary containment isolation signal upon

detection of a high main steamline tunnel differential temperature and

delete the automatic isolation function upon detection of a high main

steamline tunnel temperature. Additionally, the amendments provide a 12

hour allowed outage time for the Main Steam Line Tunnel Differential

Temperature--High isolation signal upon loss of the Reactor Building

Ventilation System.

Date of issuance: April 4, 1996.

Effective date: Immediately, to be implemented prior to restart

from refueling outage L1R07 (Unit 1) and L2R07 (Unit 2).

Amendment Nos.: 111 and 96.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the Technical Specifications.

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

FR 7281). The March 1, March 22, March 26 and April 3, 1996, submittals

provided additional clarifying information that did not change the

initial proposed no significant hazards consideration determination.

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

Safety Evaluation dated April 4, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

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

Date of application for amendments: August 25, 1995 as supplemented

on December 15, 1995, February 5, February 9, February 28, March 4,

March 28 and April 3, 1996.

Brief description of amendments: These amendments revise the

LaSalle Facility Operating Licenses and Technical Specifications (TSs)

to reflect the deletion of the leakage control system (LCS) presently

installed to control and contain the leakage past the main steamline

isolation valves (MSIVs) on each of the four main steamlines. The TSs

are also revised to raise the allowable leakage rates from 25 standard

cubic feet per hour (scfh) for each set of MSIVs and a total of 100

scfh from all four main steamlines to values of 100 scfh per steamline

and 400 scfh for all four steamlines.

Date of issuance: April 5, 1996.

Effective date: Immediately, to be implemented by startup from

refueling outage L1R07 (Unit 1) and L2R07 (Unit 2).

Amendment Nos.: 112 and 97.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the licenses and technical specifications.

Date of initial notice in Federal Register: October 25, 1995 (60 FR

54717). The December 15, 1995, February 5, February 9, February 28,

March 4, March 28 and April 3, 1996, submittals provided additional

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

April 5, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

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

Date of application for amendment: June 16, 1994, as supplemented

February 6, 1995.

Brief description of amendment: The amendment revises License

Condition 2.K and relocates the Indian Point Nuclear Generating Unit

No. 2 (IP2) fire protection requirements from the IP2 Technical

Specifications to the IP2 fire protection program plan in accordance

with the guidance provided in Generic Letter (GL) 86-10,

``Implementation of Fire Protection Requirements,'' April 24, 1986, and

GL 88-12, ``Removal of Fire Protection Requirements from Technical

Specifications,'' August 2, 1988.

Date of issuance: March 26, 1996.

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

9 months.

Amendment No.: 186.

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications and the Facility Operating License.

Date of initial notice in Federal Register: August 17, 1994 (59 FR

42335) The February 6, 1995, submittal provided clarifying information

and did not expand the scope of the original application, and 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 March 26, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

County, Michigan

Date of application for amendment: October 17, 1995.

Brief description of amendment: This amendment revises the

Palisades Facility Operating License to reference 10 CFR Part 40, allow

the use of source materials as reactor fuel, delete references to

specific amendments and specific revisions in the listed titles of the

Physical Security Plan, Suitability Training and Qualification Plan,

and the Safeguards Contingency Plan and make minor editorial changes to

the license. In addition, the Technical Specifications (TS) are

modified as follows: (1) TS 3.1.2 is modified to change the pressurizer

cooldown limit from 100 deg.F to 200 deg.F/hour; (2) the shield cooling

system requirements are relocated to the Final Safety Analysis Report;

(3) several minor editorial changes and corrections are made, including

corrections requested in the licensee's letter of March 24, 1995; and

(4) several TS bases pages have been revised. The portion of the

amendment request deleting license paragraph 2.F on reporting

requirements was denied.

Date of issuance: April 5, 1996.

Effective date: April 5, 1996.

Amendment No.: 171.

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

Facility Operating License and the Technical Specifications.

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

FR 58399).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 5, 1996, and an Environmental

Assessment dated March 11, 1996 (61 FR 10811).

[[Page 18179]]

No significant hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

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

Station, Unit No. 1, Shippingport, Pennsylvania

Date of application for amendment: December 7, 1995, as

supplemented January 4, March 1, March 5, March 7, March 11, March 27,

and March 29, 1996.

Brief description of amendment: The amendment revises Technical

Specifications 3/4.4.5 and 3/4.4.6.2 and their Bases to maintain

voltage-based steam generator tube repair criteria for the tube support

plate elevations for future cycles of operation. The amendment replaces

a 1.0 volt repair limit which had been approved on an interim basis by

License Amendment No. 184 (issued February 3, 1995) with a 2.0 volt

repair limit. The amendment also includes additional changes to reflect

the guidance provided in NRC Generic Letter 95-05, ``Voltage-Based

Repair Criteria for Westinghouse Steam Generator Tubes Affected by

Outside Diameter Stress Corrosion Cracking.''

Date of issuance: April 1, 1996.

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

within 60 days.

Amendment No: 198.

Facility Operating License No. DPR-66: Amendment revised the

Technical Specifications.

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

178) The January 4, March 1, March 5, March 7, March 11, March 27, and

March 29, 1996, letters provided clarifying information that did not

change the initial proposed no significant hazards consideration

determination or expand the amendment request beyond the scope of the

January 3, 1996 notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 1, 1996.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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: December 14, 1995.

Brief description of amendment: The amendment consists of several

changes to the instrumentation sections of the Clinton Power Station

Technical Specifications. These changes were required due to

engineering reanalyses or plant modifications. The affected

instrumentation includes: (1) steam line flow high channels for the

reactor core isolation cooling (RCIC) system, (2) ambient temperature

channels in the residual heat removal (RHR) system heat exchanger

rooms, (3) reactor vessel pressure channels that provide a permissive

for operation of the shutdown cooling mode of the RHR system, and (4)

RCIC storage tank water level instrument channels.

Date of issuance: April 10, 1996.

Effective date: April 10, 1996.

Amendment No.: 104.

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

Technical Specifications.

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

1631) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 10, 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.

No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of application for amendment: August 15, 1995, as supplemented

November 14, and December 20, 1995.

Brief description of amendment: The amendment modifies the

Monticello Technical Specifications (TS) to: (1) revise the main steam

line isolation valve leak rate test acceptance criterion to be based

upon the combined maximum flow path leakage for all four main steam

lines of 46 standard cubic feet per hour (scfh) in lieu of the current

limit of 11.5 scfh per valve; (2) revise the operability test interval

for the drywell spray header and nozzles from 5 years to 10 years; and

(3) revise TS 3/4.7.a.2, Primary Containment Integrity, to remove

information specific to the primary containment leakage rate testing

program and adopt the requirements of 10 CFR Part 50, Appendix J,

Option B, for Type A testing, while remaining under Appendix J, Option

A, for Type B and C testing.

Date of issuance: April 3, 1996.

Effective date: April 3, 1996.

Amendment No.: 95.

Facility Operating License No. DPR-22: Amendment revised the

Technical Specifications.

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

1632) The December 20, 1995, letter provided clarifying information

that was within the scope of the initial notice and did not change the

staff's initial proposed no significant hazards considerations

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

contained in a Safety Evaluation dated April 3, 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.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of application for amendment: March 1, 1996 (supersedes

December 11, 1995, application).

Brief description of amendment: The amendment modifies Technical

Specification Section 4.7, Surveillance Requirements for Primary

Containment Automatic Isolation Valves, by revising Surveillance

Requirement 4.7.D.4 to require that the seat seals of the drywell and

suppression chamber purge and vent valves be replaced every six

operating cycles.

Date of issuance: April 9, 1996.

Effective date: April 9, 1996.

Amendment No.: 96.

Facility Operating License No. DPR-22: Amendment revised the

Technical Specifications.

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

9504). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 9, 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.

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

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

Date of application for amendments: December 22, 1995.

Brief description of amendments: The amendments change Technical

Specification 3.6.1.8, ``Drywell and Suppression Chamber Purge

System,'' increasing the drywell and suppression

[[Page 18180]]

chamber purge system operating time limit from 90 hours each 365 days

to 180 hours each 365 days.

Date of issuance: March 29, 1996.

Effective date: As of date of issuance, to be implemented within 30

days.

Amendment Nos.: 115 and 77.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

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

FR 7555).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 29, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

Nuclear Power Plant, Wayne County, New York.

Date of application for amendment: February 9, 1996, as

supplemented March 20, 1996.

Brief description of amendment: The proposed amendment would revise

the Technical Specifications (TSs) to use an installed retractable

overhead door assembly and change TS 3.9.3 to satisfy closure

requirements for the containment equipment hatch during core

alterations or fuel movement in the containment building. The

retractable door is to be used as a functionally equivalent closure

plate currently required by TS 3.9.3.

Date of issuance: April 1, 1996.

Effective date: April 1, 1996.

Amendment No.: 62.

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

Technical Specifications.

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

FR 7557). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 1, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

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: August 18, 1995, as supplemented

on November 1, 1995, February 14, March 14 (there are two supplemental

letters with this date), and March 25, 1996.

Brief description of amendment: The amendment revises the Operating

License (OL) to increase the authorized core power level from 2775

Megawatts thermal (MWt) to 2900 MWt. The amendment also approves

changes to the technical specifications (TS) to implement uprated power

operation.

Date of issuance: April 12, 1996.

Effective date: April 12, 1996.

Amendment No.: 133.

Facility Operating License No. NPF-12: Amendment revises the OL and

TS.

Date of initial notice in Federal Register: December 6, 1995 (60 FR

62495). The original Federal Register notice included information from

the licensee's November 1, 1995 supplemental letter. The February 14,

March 14, and March 25, 1996 supplemental letters provided

clarification and amplification of the analysis in the November 1, 1995

letter and were not outside the scope of the initial Federal Register

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

contained in an Environmental Assessment dated April 12, 1996 and in a

Safety Evaluation dated April 12, 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 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 30, 1992, as

supplemented by letters dated September 7, 1993, August 17, 1994, and

March 7, 1996.

Brief description of amendments: These amendments add a new

technical specification (TS) 3/4.7.3.1, ``Component Cooling Water (CCW)

Safety Related Makeup System,'' and its associated Bases. The new TS

will ensure that sufficient CCW capacity is available for continued

operation of safety-related equipment during normal conditions and

design-basis events.

Date of issuance: April 11, 1996.

Effective date: April 11, 1996.

Amendment Nos.: Unit 2-129; Unit 3-118.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: March 3, 1993 (58 FR

12268). The September 7, 1993, August 17, 1994, and March 7, 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 April 11, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

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

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,

Unit 2, Hamilton County, Tennessee

Date of application for amendment: December 12, 1995, and

supplemented March 4, 1996 (TS 95-23).

Brief description of amendment: The amendment revises the TS

surveillance requirements and bases to incorporate alternate S/G tube

plugging criteria at tube support plate (TSP) intersections. The

approach taken is based on guidance given in Generic Letter (GL) 95-05,

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

Affected by Outside Diameter Stress Corrosion Cracking.'' The amendment

is applicable for Cycle 8 operation only.

Date of issuance: April 3, 1996.

Effective date: April 3, 1996.

Amendment No.: 211.

Facility Operating License Nos. DPR-77: Amendment revises the

technical specifications.

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

183) The March 6, 1996 supplemental letter provided clarifying

information which did not change the proposed no significant hazards

consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 3, 1996.

No significant hazards consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power

Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: February 27, 1996, as

supplemented by letter dated March 1, 1996.

Brief description of amendment: The amendment allows the drywell

personnel air lock shield doors to be open during Operational

Conditions 1, 2, and 3 until the end of Operating Cycle 6.

[[Page 18181]]

Date of issuance: March 22, 1996.

Effective date: March 22, 1996.

Amendment No.: 84.

Facility Operating License No. NPF-58: This amendment approved a

change to the design basis as described in the Updated Safety Analysis

Report. Public comments requested as to proposed no significant hazards

consideration: Yes (61 FR 8982 dated March 8, 1996). That notice

provided an opportunity to submit comments on the Commission's proposed

no significant hazards consideration determination. No comments have

been received. The notice also provided for an opportunity to request a

hearing BiWeekly Notice by March 18, 1996, corrected to April 5, 1996

(61 FR 10600 dated March 14, 1996), but indicated that if the

Commission makes a final no significant hazards consideration

determination any such hearing would take place after issuance of the

amendment. The March 1, 1996, supplemental letter 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 and final no

significant hazards consideration determination is contained in a

Safety Evaluation dated March 22, 1996.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

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 requests: November 21, 1995 (TXX-95288) as

supplemented by letters dated December 15, 1995 (TXX-95306), and

February 2, 1996 (TXX-96040).

Brief description of amendments: The amendments revised the core

safety limit curves and revised N-16 Overtemperature reactor trip

setpoints as a result of the reload analyses for CPSES Unit 2, Cycle 3.

In addition, the minimum required Reactor Coolant System (RCS) flow was

increased and an administrative enhancement was included in the

footnotes of the RCS flow-low reactor trip function setpoint for both

Units 1 and 2.

Date of issuance: April 1, 1996.

Effective date: April 1, 1996.

Amendment Nos.: Unit 1-49; Unit 2-35.

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

revised the Technical Specifications.

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

185) The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 1, 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, Texas 76019

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

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

Virginia

Date of application for amendments: July 26, 1995.

Brief description of amendments: The amendments revise the

Technical Specifications to increase the pressurizer safety valve lift

setpoint tolerance and reduce the pressurizer high pressure reactor

trip setpoint and allowable value.

Date of issuance: April 1, 1996.

Effective date: April 1, 1996.

Amendment Nos.: 200 and 181.

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

the Technical Specifications.

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

45189) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 1, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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

Generating Station, Coffey County, Kansas

Date of amendment request: March 8, 1996, as supplemented by letter

dated March 26, 1996.

Brief description of amendment: This amendment reduces the

calculated thermal design flow of the reactor coolant system and

increases the trip setpoint of the low pressurizer pressure.

Date of issuance: April 4, 1996.

Effective date: April 4, 1996.

Amendment No.: 99.

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

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: Yes (61 FR 10389 dated March 13, 1996). The notice

provided an opportunity to submit comments on the Commission's proposed

no significant hazards consideration determination. No comments have

been received. The notice also provided for an opportunity to request a

hearing by April 12, 1996, but indicated that if the Commission makes a

final no significant hazards consideration determination any such

hearing would take place after issuance of the amendment.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 4, 1996.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

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

66621.

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.

[[Page 18182]]

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 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 May 24, 1996, the licensee

may file a request for a hearing with respect to issuance of the

amendment to the subject facility operating license and any person

whose interest may be affected by this proceeding and who wishes to

participate as a party in the proceeding must file a written request

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

hearing and a petition for leave to intervene shall be filed in

accordance with the Commission's ``Rules of Practice for Domestic

Licensing Proceedings'' in 10 CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714 which is available at the

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

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

particular facility involved. If a request for a hearing or petition

for leave to intervene is filed by the above date, the Commission or an

Atomic Safety and Licensing Board, designated by the Commission or by

the Chairman of the Atomic Safety and Licensing Board Panel, will rule

on the request and/or petition; and the Secretary or the designated

Atomic Safety and Licensing Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) The nature of the petitioner's right under the

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

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

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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, 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

[[Page 18183]]

General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC

20555, 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 the

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

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo

Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

Date of application for amendment: April 1, 1996, as supplemented

by letter dated April 3, 1996.

Brief description of amendment: The amendment modifies Technical

Specification (TS) 3/4.9.6 to temporarily allow the use of a hoist

instead of the refueling machine for the movement of the fuel assembly

at core location A-07.

Date of issuance: April 3, 1996.

Effective date: April 3, 1996.

Amendment No.: Unit 2--96.

Facility Operating License No. NPF-51: The amendment revised the

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: No.

The Commission's related evaluation of the amendment, finding of

emergency circumstances, and final determination of no significant

hazards consideration are contained in a Safety Evaluation dated April

3, 1996.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: William H. Bateman.

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 application of amendments: April 2, 1996.

Brief description of amendments: The amendments revise Technical

Specification (TS) Section 4.5.4, ``Penetration Room Ventilation

System'' and TS Section 4.14, ``Reactor Building Purge Filters and

Spent Fuel Pool Ventilation System.'' The change updates the industry

guidance reference for testing charcoal absorber units for the system

covered by those TS.

Date of Issuance: April 2, 1996.

Effective date: April 2, 1996, to be implemented within 30 days.

Amendment Nos.: 215, 215, and 212.

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The

amendments revised the Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: No.

The Commission's related evaluation of the amendments, finding of

emergency circumstances, and final determination of no significant

hazards consideration are contained in a Safety Evaluation dated April

2, 1996.

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.

Dated at Rockville, Maryland, this 17th day of April 1996.

For the Nuclear Regulatory Commission.

Steven A. Varga,

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

Regulation.

[FR Doc. 96-9925 Filed 4-23-96; 8:45 am]

BILLING CODE 7590-01-P