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 September 29, 1995, through October 13,

1995. The last biweekly notice was published on October 11, 1995 (60 FR

52927).

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 November 24, 1995, 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

[[Page 54715]]

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 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.

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

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

1, 2, and 3, Maricopa County, Arizona

Date of amendments request: June 13, 1995, as supplemented by

letter dated August 16, 1995.

Description of amendments request: The proposed amendments would

extend allowed outage times (AOTs) for a safety injection tank (SIT), a

low- pressure safety injection (LPSI) subtrain, and an emergency diesel

generator (EDG) and add the bases for the extended AOTs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The Safety Injection Tanks (SITs) are passive components in the

Emergency Core Cooling System. The SITs are not an accident

initiator in any accident previously evaluated. Therefore, this

change does not involve a significant increase in the probability of

an accident previously evaluated.

SITs were designed to mitigate the consequences of Loss of

Coolant Accidents (LOCA). These proposed changes do not affect any

of the assumptions used in deterministic LOCA analysis. Hence the

consequences of accidents previously evaluated do not significantly

increase.

The allowed outage time (AOT) extension for boron concentration

outside the prescribed limits does not involve a significant

increase in the consequences of an accident as evaluated and

approved by the NRC in NUREG-1432, ``Standard Technical

Specifications for Combustion Engineering Plants.'' These changes

are applicable to PVNGS.

The changes pertaining to SIT inoperability based solely on

instrumentation malfunction do not involve a significant increase in

the consequences of an accident as evaluated and endorsed by the NRC

in NUREG-1366, ``Improvements to Technical Specifications

Surveillance Requirements,'' and Generic Letter 93-05, ``Line-Item

Technical Specifications Improvements to Reduce Surveillance

Requirements for Testing During Power Operations.'' These changes

are applicable to PVNGS.

The AOT extension from one hour to 24 hours for a SIT that is

inoperable due to reasons other than boron concentration not within

limits or the inability to verify level or pressure does not involve

a significant increase in the consequences of an accident. In order

to fully evaluate the affect of the SIT AOT extension, probabilistic

safety analysis (PSA) methods were utilized. The results of these

analyses show no significant increase in the core damage frequencies

(CDF). As a result, there would be no significant increase in the

consequences of an accident previously evaluated. These analyses are

detailed in CE NPSD-994, Combustion Engineering Owners Group ``Joint

Applications Report for Safety Injection Tank AOT/STI Extension,''

May 1995.

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

or different kind of accident from any accident previously

evaluated.

This proposed change does not change the design, configuration,

or method of operation of the plant. Therefore, this change does not

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

any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not involve a significant reduction in a

margin of safety.

The proposed changes do not affect the limiting conditions for

operation or their bases that are used in the deterministic analyses

to establish the margin of safety. PSA evaluations were used to

evaluate these changes. These evaluations demonstrated that the

changes are either risk neutral or risk beneficial. These

evaluations are detailed in CE NPSD-994.

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

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

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

amendments request involve no significant hazards consideration.

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.

[[Page 54716]]

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

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

Date of amendment request: September 11, 1995.

Description of amendment request: The proposed change is to (1)

modify a limiting condition for operation (LCO), TS Section 3.10.1.3,

to provide for temporary conditions in which the full length control

rod insertion limits (RILs) are exceeded due to automatic plant

responses or conservative operator actions and (2) add an allowance for

RILs to be exceeded for a time no greater than the time criteria

established by the axial power distribution methodology or 1 hour,

whichever is sooner. An action is added for the reactor to be placed in

the hot shutdown condition within 6 hours if compliance with the RILs

cannot be restored within the specified time period.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

This proposed change does not involve a significant hazards

consideration for the following reasons.

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed change does not involve the addition or

modification of plant equipment, nor does it alter the design,

material, or operation of plant systems. No analyzed accidents are

initiated by an entire control rod bank exceeding the RILs, due to

automatic plant responses or conservative operator actions. The

overall performance of the Reactor Control System, Power

Distribution Control procedures, and Control Rod Drive System is not

degraded. There is no increase in fatigue or number of operational

cycles of equipment, and there is no change in system interfaces.

The consequences of previously evaluated accidents are not increased

since exceeding the RILs for a limited period is acceptable as the

probability of a simultaneous occurrence of an independent accident

is low. Therefore, an allowance for RILs to be exceeded for a

maximum of one (1) hour does not affect the probability of

occurrence or consequences of an analyzed accident.

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

or different kind of accident from any accident previously

evaluated. The proposed change adds an allowance for RILs to be

exceeded for a maximum of one (1) hour. The proposed change does not

involve the addition or modification of plant equipment, nor does it

alter the design or operation of plant systems. The only procedural

changes required will be those associated with recovery from the

infrequent condition of exceeding the RILs. No new accident

scenarios are introduced when the RILs are exceeded for a short

period of time due to automatic plant responses or conservative

operator actions because the probability of a simultaneous

occurrence of an independent accident is low. Therefore, an

allowance for RILs to be exceeded for a maximum of one (1) hour does

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

from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety. The proposed change adds an allowance for

RILs to be exceeded for a maximum of one (1) hour. The proposed

change does not involve the addition or modification of plant

equipment, nor does it alter the design or operation of plant

systems. The overall performance of the Reactor Control System,

Power Distribution Control, and Control Rod Drive System is not

degraded. There is no increase in fatigue or number of operational

cycles of equipment, and there is no change in system interfaces.

When the RILs are exceeded for a limited time period, due to

automatic plant responses or conservative operator actions, the

margin of safety is not reduced because the probability of a

simultaneous occurrence of an independent accident is acceptably

low. Therefore, an allowance for RILs to be exceeded for a maximum

of one (1) hour does not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

Attorney for licensee: R.E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Project Director: David B. Matthews.

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

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

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

and 2, Will County, Illinois

Date of amendment request: September 14, 1995.

Description of amendment request: The proposed amendment would

allow the use of an alternate zirconium based fuel cladding, ZIRLO, and

permit limited substitution of ZIRLO filler rods for fuel rods. The

proposed amendment also includes a clarification and an editorial

change.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The methodologies used in the accident analyses remain

unchanged. The proposed changes do not change or alter the design

assumptions for the systems or components used to mitigate the

consequences of an accident. Use of ZIRLO fuel cladding does not

adversely affect fuel performance or impact nuclear design

methodology. Therefore, accident analysis results are not impacted.

The operating limits will not be changed and the analysis

methods to demonstrate operation within the limits will remain in

accordance with NRC-approved methodologies. Other than the changes

to the fuel assemblies, there are no physical changes to the plant

associated with this Technical Specification change. A safety

analysis will continue to be performed for each cycle to demonstrate

compliance with all fuel safety design bases.

VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods meet the

same fuel assembly and fuel rod design bases as other VANTAGE 5 fuel

assemblies. In addition, the 10 CFR 50.46 criteria are applied to

the ZIRLO clad fuel rods. The use of these fuel assemblies will not

result in a change to the reload design and safety analysis limits.

Since the original design criteria are met, the ZIRLO clad fuel rods

will not be an initiator for any new accident. The clad material is

similar in chemical composition and has similar physical and

mechanical properties as Zircaloy-4. Thus, the cladding integrity is

maintained and the structural integrity of the fuel assembly is not

affected. ZIRLO cladding improves corrosion performance and

dimensional stability. No concerns have been identified with respect

to the use of an assembly containing a combination of Zircaloy-4 and

ZIRLO clad fuel rods. Since the dose predictions in the safety

analyses are not sensitive to the fuel rod cladding material used,

the radiological consequences of accidents previously evaluated in

the safety analysis remain valid.

Replacing the reference to the Final Safety Analysis Report

(FSAR) with a reference to the Updated Final Safety Analysis Report

(UFSAR) is an editorial change to reflect the current document.

Adding that reload fuel shall be similar in physical design to the

initial core loading or previous cycle loading is a clarification. A

reload analysis is completed for each cycle, in accordance with

USNRC-approved methodologies.

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.

[[Page 54717]]

VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods satisfy the

same design bases as those used for other VANTAGE 5 fuel assemblies.

All design and performance criteria continue to be met and no new

failure mechanisms have been identified. The ZIRLO cladding material

offers improved corrosion resistance and structural integrity.

The proposed changes do not affect the design or operation of

any system or component in the plant. The safety functions of the

related structures, systems, or components are not changed in any

manner, nor is the reliability of any structure, system, or

component reduced. The changes do not affect the manner by which the

facility is operated and do not change any facility design feature,

structure, or system. No new or different type of equipment will be

installed. Since there is no change to the facility or operating

procedures, and the safety functions and reliability of structures,

systems, or components are not affected, 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 use of Zircaloy-4, ZIRLO, or stainless steel filler rods in

fuel assemblies will not involve a significant reduction in the

margin of safety because analyses using NRC-approved methodology

will be performed for each configuration to demonstrate continued

operation within the limits that assure acceptable plant response to

accidents and transients. These analyses will be performed using

NRC-approved methods that have been approved for application to the

fuel configuration.

Use of ZIRLO cladding material does not change the VANTAGE 5

reload design and safety analysis limits. The use of these fuel

assemblies will take into consideration the normal core operating

conditions allowed in the Technical Specifications. For each cycle

reload core, the fuel assemblies will be evaluated using NRC-

approved reload design methods, including consideration of the core

physics analysis peaking factors and core average linear heat rate

effects.

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

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

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

requested amendments involve no significant hazards consideration.

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

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

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

Wilmington, Illinois 60481

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

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

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

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

Date of amendment request: August 28, 1995.

Description of amendment request: The proposed amendments would

support elimination of the Main Steam Isolation Valve Leakage Control

System (MSIV LCS) and instead use the main steamline drains and

condenser to process MSIV leakage. The proposed changes would also

increase the allowable MSIV leakage from 100 standard cubic feet per

hour (scfh) for all four main steam lines to 100 scfh per steam line

(400 scfh for all four main steam lines).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because:

The proposed changes involve eliminating the requirement for the

Main Steam Isolation Valve Leakage Control System (MSIV LCS). This

system is manually initiated following a design basis Loss of

Coolant Accident (LOCA). Since operation of the LCS is initiated

after the accident has already begun, elimination of that system

will not affect the probability of a LOCA. The LCS only interfaces

with the main steamlines, with the exception of one MSIV LCS power

supply which supplies power to the Reactor Protection System Scram

Discharge Volume high level scram. This power supply will remain in

place after the MSIV LCS is isolated from the main steamlines.

Therefore, since the only significant system interface is with the

main steamlines, and the system does not impact the reliability of

any plant equipment, elimination of that system will not cause an

increase in the likelihood that any accident might occur.

The proposed change to increase the allowable MSIV leakage limit

from 100 scfh through all four main steam lines to 100 scfh per main

steam line (400 scfh total) will not increase the probability of an

accident. MSIV operability will not be degraded with the allowed

increased leakage.

The consequences of a LOCA are not significantly increased and

do not exceed the previously accepted licensing criteria for this

accident. General Electric has calculated the revised LOCA doses,

which have been added to the previous LOCA doses. These resulting

values are well below the acceptance criteria of 10CFR100 and

10CFR50, Appendix A.

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

different kind of accident from any accident previously evaluated

because:

The proposed changes require the use of the main steam piping

and condenser to process MSIV leakage. The analyses presented

provide assurance that this additional function does not compromise

the reliability of those systems. They will therefore continue to

function as intended and not be subject to an increased failure rate

or a failure of a different kind than previously considered.

In addition, MSIV functionality will not be adversely impacted

as a result of the increased leakage limit. The MSIVs are not being

modified in any way and will continue to provide their intended

isolation function.

The MSIV LCS will be cut and capped, which will completely

isolate it from other plant systems. Future degradation of its

associated piping would not impact any other system or create a

failure not previously analyzed. However, piping seismic Class II

over I criteria must be maintained for the abandoned MSIV LCS piping

until it is removed from the plant.

The proposed changes do not involve a significant reduction in a

margin of safety because:

The proposed change has been evaluated with respect to dose

limits contained in 10CFR100 and 10CFR50, Appendix A. The revised

dose calculations verify that the use of the main steam lines and

the condenser for leakage control, in place of the MSIV LCS, and

with an allowable total leakage of 400 scfh, maintains adequate

margins to the criteria listed above.

Even though there is a reduction in the margin to safety, the

new doses remain well within the criteria of 10 CFR 100 and 10 CFR

50, Appendix A. This reduction in margin is not significant when

compared to the increased reliability and capability of the main

steam lines and condenser as a method of treating MSIV leakage. The

new leakage pathway is consistent with the philosophy of protection

by multiple barriers for limiting fission product release to the

environment. In addition, the new method is passive and does not

require any new logic control or interlocks. The new pathway is also

capable of handling a larger amount of leakage than the MSIV LCS,

which was previously subject to concerns that it would not function

at leakage rates higher than its design capacity, or at reactor

pressures greater than 35 psig.

The revised calculated LOCA doses remain well within the

regulatory limits for MSIV leakage rates of 400 scfh for all four

main steam lines (100 scfh per steam line), and the margin to safety

is not significantly reduced as a result of the proposed changes.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

Austin, One First National Plaza, Chicago, Illinois 60603.

[[Page 54718]]

NRC Project Director: Robert A. Capra.

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: September 1, 1995.

Description of amendment request: Generic Letter 88-16 provided

guidance on removing cycle-specific parameters which are calculated

using NRC-approved methodologies from the Technical Specifications

(TS). The parameters are replaced in the TS with a reference to a named

report which contains the parameters, and a requirement that the

parameters remain within the limits specified in the report. The

proposed changes incorporate NRC-approved methodologies, approved

revisions to previously approved methodologies, or republished versions

of previously approved methodologies into section 6.9.2 of the Oconee

TS. The limits to which these methodologies are applied are (1) Axial

Power Imbalance Protective Limits and Variable Low RCS Pressure

Protective Limits, (2) Reactor Protective System Trip Setting Limits

for the Flux/Flow/Imbalance and Variable Low Reactor Coolant System

Pressure Trip Functions, and (3) Power Imbalance Limits. Since the

proposed changes only incorporate NRC-approved methodologies into the

TS, the licensee proposed that the changes are administrative in nature

and can be assumed to have no impact, or potential impact, on the

health and safety of the public.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes will not create a significant hazards

consideration, as defined by 10 CRF 50.92, because:

(1) The proposed changes will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes are administrative in nature, and do not

affect any system, procedure, or manipulation of any equipment which

could affect the probability or consequences of any accident.

(2) The proposed changes will not create the possibility of any

new or different kind of accident from any accident previously

evaluated.

The proposed changes are administrative in nature, and cannot

introduce any new failure mode or transient which could create any

accident.

(3) The proposed changes will not involve a significant

reduction in a margin of safety.

The proposed changes are administrative in nature, and will not

affect any operating parameters or limits which could result in a

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: September 25, 1995.

Description of amendment request: The proposed amendment adds a

repair limit for circumferential cracks in steam generator tubes. It

deletes the requirement to repair cracks that are within the repair

limit. The proposed amendment also reduces the primary-to-secondary

leak rate limit.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

Consistent with draft Regulatory Guide (RG) 1.121, ``Basis for

Plugging Degraded PWR Steam Generator Tubes,'' the traditional

maximum depth based criteria for steam generator tube repair

implicitly ensures that tubes accepted for continued service will

retain adequate structural and leakage integrity during normal

operating, transient, and postulated accident conditions. It is

recognized that defects in tubes permitted to remain in service

occasionally grow through-wall and develop small leaks. Limits on

allowable primary-to-secondary leakage established in the technical

specifications ensure timely plant shutdown before the structural

and leakage integrity of the affected tube is challenged.

The proposed change to implement a circumferential crack repair

limit in the expansion transition region for ANO-2 meets the

criteria of RG 1.121. The 40% degraded area repair limit was

determined by performing a structural analysis per the

recommendations of the RG and applying the following uncertainties:

95% lower bound material properties, 95% lower bound burst curve,

95% lower bound eddy current measurement uncertainties, and 95%

upper bound crack growth rate. The analysis demonstrates that tube

leakage and conditional probability of burst are acceptably low

during either normal operation or the most limiting accident

condition, a postulated main steam line break (MSLB) event.

As part of the implementation of the circumferential crack

repair limit, the distribution of End-of-Cycle (EOC) circumferential

indications in the expansion transition region will be used to

calculate the primary-to-secondary leakage. The allowable leakage is

bounded by the maximum leakage which results in doses within the

applicable dose limits (10CFR100 and General Design Criteria 19).

The limit is calculated using the technical specification reactor

coolant system (RCS) iodine activity. Application of the

circumferential crack repair limit requires the projection of the

postulated MSLB leakage based on the projected EOC distribution for

the next cycle. The projected EOC distribution is developed using

the most recent EOC eddy current results based on crack arc length.

The reduction in the leak rate limit reduces the possibility

that a defect in a leaking tube will grow to a size that is not

structurally acceptable.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

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

Different Kind of Accident from any Previously Evaluated.

Implementation of the proposed circumferential crack repair

limit does not introduce any significant changes to the plant design

basis. The only accident possible from implementation of this limit

is a tube rupture, which has already been evaluated in the ANO-2

Safety Analysis Report.

The maximum primary-to-secondary leakage rate has been reduced

to 150 gallons per day through any one steam generator to help

preclude the potential for excessive leakage during all plant

conditions. The RG 1.121 criterion for establishing the operational

leak rate limit considers: (1) the detection of a crack before

potential tube rupture as a result of faulted plant conditions; (2)

the maintenance of a margin to tube rupture of not less than three

for normal operating conditions; and (3) that any leakage rate

increase will be gradual to provide time for corrective action. The

150 gallon per day limit is intended to provide for leakage

detection and plant shutdown in the event of an unexpected crack

propagation resulting in excessive leakage.

Steam generator tube integrity is maintained through inservice

inspection and primary-to-secondary leakage monitoring. Any tubes

exceeding the circumferential crack repair limit are removed from

service.

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

or different kind of accident from any previously evaluated.

Criterion 3--Does not Involve a Significant Reduction in the

Margin of Safety.

The use of the circumferential crack repair limit will maintain

steam generator tube

[[Page 54719]]

integrity commensurate with the criteria of RG 1.121. Upon

implementation of the limit, even under worst case conditions, the

occurrence of circumferential cracking in the expansion transition

region is not expected to lead to a steam generator tube rupture

event during normal or faulted plant conditions. The distribution of

crack indications left in service will result in acceptable primary-

to-secondary leakage and conditional tube burst probability during

all plant conditions.

The installation of steam generator tube plugs and sleeves

reduces RCS flow margin. Implementation of the circumferential crack

repair limit will decrease the number of tubes which must be

repaired by plugging or sleeving, thereby retaining additional flow

margin that would otherwise be reduced.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

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

NRC Project Director: William D. Beckner.

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

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: May 5, 1995, as supplemented September

28, 1995.

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications (TS) by revising TS

2.1.1, Safety Limit--Reactor Core; TS 2.2, Limiting Safety System

Settings--Reactor Trip System Instrumentation Setpoints; TS 3/4.2.5

Power Distribution Limits-- Departure from Nucleate Boiling (DNB)

Parameters; TS 3/4.3.2 Engineered Safety Features Actuation System

Instrumentation and the associated BASES. The proposed revision to the

TS includes (a) the implementation of Westinghouse's NRC approved

Revised Thermal Design Procedure (RTDP), and (b) a revision to the

Steam Generator Water Level Low-Low trip setpoint.

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 licensee's analysis was

presented separately for the following areas: core thermal limits,

overtemperature [delta] T and overpower [delta] T reactor trip

setpoint; steam generator process measurement accuracy; and DNB

parameter surveillance requirements.

Core Thermal Limits, overtemperature [delta] T and overpower [delta] T

Reactor Trip Setpoint

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The revised Overtemperature and Overpower [delta] T reactor trip

functions do not involve an increase in the probability or

consequences of an accident previously evaluated because operation

with these revised values will not cause any design or analysis

acceptance criteria to be exceeded. The structural and functional

integrity of all plant systems is unaffected. The Overtemperature

and Overpower [delta] T reactor trip functions are part of the

accident mitigation response and are not initiators for any

transient. Therefore, the probability of occurrence previously

evaluated are not affected.

The changes to the Overtemperature and Overpower [delta] T

reactor trip functions do not affect the integrity of the fission

product barriers utilized for mitigation of radiological dose

consequences as a result of an accident. In addition, the off-site

mass releases used as input to the dose calculations are unchanged

from those previously assumed. Therefore, the off-site dose

predictions remain within the acceptance criteria of 10 CFR Part 100

limits for each of the transients affected. Since it has been

concluded that the transient analyses results are unaffected by the

parameter modifications, it is concluded that the probability or

consequences of an accident previously evaluated are not increased.

(2) The proposed license amendments do not create the

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

previously evaluated.

The revised Overtemperature and Overpower [delta] T reactor trip

functions do not create the possibility of a new or different kind

of accident from any accident previously evaluated because the

setpoint adjustments do not affect accident initiation sequences. No

new operating configuration is being imposed by the setpoint

adjustments that would create a new failure scenario. In addition,

no new failure modes or limiting single failures have been

identified. Therefore, the types of accidents defined in the UFSAR

continue to represent the credible spectrum of events to be analyzed

which determine safe plant operation. Therefore, it is concluded

that no new or different kind of accidents from those previously

evaluated have been created as a result of these revisions.

(3) The proposed license amendments do not involve a significant

reduction in a margin of safety.

The changes to the Overtemperature and Overpower [delta] T

reactor trip functions do not involve a reduction in the margin of

safety because the margin of safety associated with the

Overtemperature and Overpower [delta] T reactor trip functions, as

verified by the results of the accident analyses, are within

acceptable limits. All transients impacted by implementation of the

RTDP methodology have been analyzed and have met the applicable

accident analyses acceptance criteria. The margin of safety required

for each affected safety analysis is maintained. This conclusion is

not changed by the Overtemperature and Overpower [delta] T setpoint

modifications. The adequacy of the revised Technical Specifications

values to maintain the plant in a safe operating condition has been

confirmed. Therefore, the changes to the Overtemperature and

Overpower [delta] T reactor trip functions do not involve a

significant reduction in the margin of safety.

Steam Generator Process Measurement Accuracy

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The revised reactor trip setpoints on Steam Generator water

level do not involve a significant increase in the probability or

consequences of an accident previously evaluated. Operation with

these revised values will not cause any design or analysis

acceptance criteria to be exceeded. The structural and functional

integrity of any plant system is unaffected. The Steam Generator

Water Level trip functions are part of the accident mitigation

response and are not themselves initiators for any transient.

Therefore, the probability of occurrence previously evaluated is not

affected.

The changes to the reactor trip setpoints do not affect the

integrity of the fission product barriers utilized for mitigation of

radiological dose consequences as a result of an accident. The Steam

Generator Water Level Low-Low trip setpoint assumed in the safety

analyses has been revised and acceptable results were obtained. The

Steam Generator Water Level-Low setpoint is not credited in the

safety analysis. Consequently, the required margin of safety for

each affected safety analysis has been maintained. In addition, the

offsite mass releases used as input to the dose calculations are

unchanged from those previously assumed. Therefore, the offsite dose

predictions remain within the acceptance criteria of 10 CFR Part 100

limits for each of the transient analyses affected. Since it has

been determined that the transient analysis results are unaffected

by these parameter modifications, FPL concludes that the

consequences of an accident previously evaluated are not increased.

(2) The proposed license amendments do not create the

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

previously evaluated.

The setpoint values do not affect the assumed accident

initiation sequences. In addition, no new failure modes or limiting

single failures have been identified for any

[[Page 54720]]

plant equipment. Therefore, the types of accidents defined in the UFSAR

continue to represent the credible spectrum of events to be analyzed

which determine safe plant operation. Therefore, the possibility of

a new or different kind of accident from any accident evaluated is

not increased.

(3) The proposed license amendments do not involve a significant

reduction in the margin to safety.

The current Technical Specification trip setpoints and allowable

values were changed to maintain the current safety analysis limits.

The Steam Generator Water Level Low-Low trip setpoint assumed in the

safety analyses has been revised and acceptable results were

obtained. The Steam Generator Water Level-Low setpoint is not

credited in the safety analysis. Consequently, the required margin

of safety for each affected safety analysis has been maintained.

Thereby, the adequacy of the revised Technical Specification values

to maintain the plant in a safe operating condition is also

confirmed.

DNB Parameter Surveillance Requirements

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

With the retention of the previous Safety Analyses Limits for

Departure from Nucleate Boiling (DNB) (T.S. 3/4.2.5) and the

existing Reactor Coolant System (RCS) low flow trip Nominal Trip

Setpoint (NTS), there is no increase in the probability or

consequences of an accident previously evaluated because there is no

change to any design or analysis acceptance criteria. The structural

and functional integrity of any plant system is unaffected. The

proposed license amendments revise the surveillance requirements for

DNB parameters and incorporate the RTDP uncertainty analysis into

the Westinghouse methodology for the RCS Loss of Flow determination

of the Allowable Value.

The changes to the reactor trip functions do not affect the

integrity of the fission product barriers utilized for mitigation of

radiological dose consequences as a result of an accident. The

margin to safety for the RCS Loss of Flow trip remains protected as

the trip setpoints assumed in the safety analyses are not revised.

In addition, the offsite mass releases used as input to the dose

calculations are unchanged from those previously assumed. Therefore,

the offsite dose predictions remain within the acceptance criteria

of 10 CFR Part 100 limits for each of the transients affected. Since

it has been determined that the transient results are unaffected by

these parameter modifications, it is concluded that the consequences

of an accident previously evaluated are not increased.

(2) Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The revised Allowable Value does not create the possibility of a

new or different kind of accident from any accident previously

evaluated. Revision of the surveillance requirements merely provides

clarification to more accurately reflect the surveillance activity.

The Allowable Value does not affect the assumed accident

initiation sequences. In addition, no new failure modes or single

failures have been identified for any plant equipment. Therefore,

the types of accidents defined in the UFSAR continue to represent

the credible spectrum of events to be analyzed which determine safe

plant operation. Therefore, it is concluded that no new or different

kind of accidents from those previously evaluated have been created

as a result of these revisions.

(3) The proposed license amendments do not involve a significant

reduction in the margin to safety.

The RCS Loss of Flow setpoint assumed in the safety analysis

remains unchanged. Since the safety analysis limit setpoint value is

unchanged and no safety analysis is affected, the required margin of

safety for each affected safety analysis is maintained. Thereby, the

adequacy of the revised Technical Specification values to maintain

the plant in a safe operating condition is also confirmed.

Therefore, the change to the RCS Loss of Flow Allowable Value 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 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: Florida International

University, University Park, Miami, Florida 33199.

Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Project Director: David B. Matthews.

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

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: July 26, 1995, as supplemented by letter

dated October 4, 1995.

Description of amendment request: The licensee proposes to revise

the technical specifications surveillance intervals and allowed outage

times for the channel operational tests performed on the analog

``bistable'' comparator modules for the reactor trip, reactor trip

permissive functions, engineered safety features actuation and

permissive functions identified below.

TS Table 3.3-1--Revise ACTION Statements 2a, 6, 12 and 13; increase

the time allowed for a channel to be inoperable or out of service in an

untripped condition from 1 hour to 6 hours. Revise ACTION Statement 2b;

increase the time a Nuclear Instrumentation System (NIS) channel in a

functional group may be bypassed to perform testing from 2 to 4 hours.

TS Table 3.3-2--Revise ACTION Statement 14; increase the time to be

in HOT STANDBY with the number of OPERABLE channels one less that the

Minimum Channels OPERABLE requirement from 6 to 12 hours. Revise ACTION

Statements 14, 20 and 22; increase the allowed outage time for test of

the logic trains from 2 hours to 8 hours. Revise ACTION Statements 15,

18 and 25; increase the time allowed for a channel to be inoperable and

out of service in an untripped condition from 1 hour to 6 hours.

TS Table 4.3-1--Revise the surveillance interval for Items 2.a, 4,

7, 8, 10, 11, 12 and Note (9) from monthly to quarterly. Revise the

surveillance interval for Item 2.b from monthly to startup, and Item 3

from monthly/startup to startup only. Revise the surveillance interval

for Items 17.a, 17.b, 17.c and 17.d from monthly to refueling. Revise

Note (1) from ``7 days'' to ``31 days'' and delete Note (8).

TS Table 4.3-2--Revise the surveillance interval for Items 1.d,

1.e, 1.f, 4.d, 5.c, 6.b, and 8.a from monthly to quarterly.

TS BASES 3/4.3.1 and 3/4.3.2--Revise the BASES section for

Technical Specification Sections 3/4.3.1 and 3/4.3.2 to reference the

Westinghouse WCAPs 10271 and 10271, Supplement 2, and associated

Nuclear Regulatory Commission (NRC) safety evaluation reports (SERs).

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) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes in Technical Specification surveillance

intervals and allowed outage times for the subject Reactor

Protection System (RPS)/Nuclear Instrumentation System (NIS)/

Engineered Safety Features Actuation System (ESFAS) analog

instrumentation have been revised in accordance with the

recommendations and criteria of Westinghouse WCAP-10271, WCAP 10271,

Supplement 2, and the NRC's SERs on the same subject dated February

21, 1985 and dated February 22, 1989.

The proposed changes do not involve any hardware or setpoint

changes. Similarly, the proposed changes do not alter the manner in

which safety limits, limiting safety system setpoints or limiting

conditions for operation

[[Page 54721]]

are determined. Implementation of the proposed changes does affect the

probability of failure of the RPS, including NIS, and ESFAS, but

does not alter the manner in which protection is afforded nor the

manner in which limiting setpoint criteria are established for the

RPS/ESFAS instrumentation systems. Consequently, the proposed

changes do not result in an increase in the severity or consequences

of any accident previously evaluated.

Implementation of the proposed changes is expected to result in

an acceptably small increase in total RPS unavailability. This

increase is primarily due to less frequent surveillances and was

generically quantified to be less than 3% within WCAP-10271. WCAP-

10271 also documents that the implementation of the proposed changes

is also expected to result in a significant reduction in the

probability of core melt from inadvertent reactor trips (WCAP-

10271). This is the result of a reduction in the number of

inadvertent reactor trips (0.5 fewer inadvertent reactor trips per

unit per year) occurring during testing of the RPS instrumentation.

This reduction is primarily attributable to testing in bypass for

applicable channels and to less frequent surveillances. WCAP-10271

documents that the reduction in inadvertent core melt probability is

sufficiently large to counter the increased core melt probability,

resulting in an overall reduction in total core melt probability of

approximately 1%.

A corresponding probabilistic risk assessment (WCAP-10271,

Supplement 2) was documented by Westinghouse for the generic

implementation of the proposed changes for ESFAS instrumentation.

This Westinghouse evaluation along with the independent assessments

performed by an NRC contractor demonstrated that a 6% core damage

frequency increase represented an upper bound for Westinghouse

plants. For more realistic testing strategies, the core damage

frequency increase would be substantially less than this.

Consequently, the changes in Technical Specifications associated

with an extension of the surveillance intervals and out of service

times for the RPS/ESFAS instrumentation systems will have only a

small impact on plant risk. On this basis, FPL concludes that the

proposed changes will not have a significant effect on the

probability or consequences of licensing basis events; and the

probability or consequences of an accident previously evaluated for

Turkey Point does not significantly increase.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed changes in Technical Specification surveillance

intervals and allowed outage times for the subject RPS/ESFAS analog

instrumentation have been revised in accordance with the

recommendations and criteria of Westinghouse WCAP-10271, WCAP 10271,

Supplement 2, and the NRC's SERs on the same subject dated February

21, 1985 and dated February 22, 1989.

The proposed changes do not involve any hardware or setpoint

changes. Some existing instrumentation is designed to be tested in

bypass and current Technical Specifications allow testing in bypass.

Testing in bypass is also recognized by IEEE Standards.

Therefore, testing in bypass has been previously approved and

implementation of the proposed changes for testing in bypass does

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

from any previously evaluated. Furthermore, since the proposed

changes do not alter the manner in which protection is afforded nor

the manner in which limiting criteria are established for the RPS

and ESFAS instrumentation systems, the possibility of a new or

different kind of accident from any previously evaluated has not

been created.

The proposed changes do not result in a change in the manner in

which the RPS or ESFAS provides plant protection. No change is being

made which alters the function of the RPS or ESFAS (other than in a

test mode). Rather, the likelihood or probability of the RPS and

ESFAS functioning properly is the only effect.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident nor involve a reduction in a

margin of safety as defined in the Safety Analysis Report.

Consequently, the changes in Technical Specifications associated

with an extension of the surveillance intervals and out of service

times for the RPS/ESFAS instrumentation systems will not create the

possibility of a new or different kind of accident from any

previously evaluated by the NRC, and does not create the possibility

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

evaluated.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes in Technical Specification surveillance

intervals and allowed outage times for the subject RPS/ESFAS analog

instrumentation have been revised in accordance with the

recommendations and criteria of Westinghouse WCAP-10271, WCAP 10271,

Supplement 2, and the NRC's SERs on the same subject dated February

21, 1985 and dated February 22, 1989.

These changes in Technical Specifications only affect the

frequency of the channel operational tests and the allowed outage

times; they do not alter the manner in which protection is afforded

nor the manner in which limiting setpoint criteria are established.

In addition, the fundamental process to implement these channel

operational tests remains the same.

The proposed changes do not alter the manner in which safety

limits, limiting safety system setpoints or limiting conditions for

operation are determined. The impact of reduced testing is to allow

a longer time interval over which instrument uncertainties (e.g.,

drift) may act. The site specific review of historical drift data

and the conservative application of drift in the Westinghouse

methodology are sufficient to demonstrate that the basis of the

Technical Specification setpoint determinations are not adversely

affected by extending the surveillance interval from monthly to

quarterly, that is, quarterly surveillance test intervals would not

exceed the allowable instrument drift of these analog devices.

Implementation of the proposed changes is expected to result in

an overall improvement in safety by:

(a) Fewer inadvertent reactor trips per unit per year. This is

due to less frequent testing which minimizes the time spent in a

partial trip condition.

(b) Higher quality repairs leading to improved equipment

reliability due to longer allowed repair times.

(c) Improvements in the effectiveness of the operating staff in

monitoring and controlling plant operation. This is due to less

frequent distractions of the operator and shift supervisor from

attending to instrumentation testing.

The Westinghouse analysis demonstrates that any expected

increases in probability of core melt or core damage frequency are

small and are therefore acceptable. Consequently, the changes in

Technical Specifications associated with an extension of the

surveillance intervals and out of service times for the RPS/ESFAS

instrumentation systems will not significantly reduce the margin of

plant safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Attorney for licensee: J.R. Newman, Esquire, Morgan, Lewis &

Bockius, 1800 M Street NW., Washington, DC 20036.

NRC Project Director: David B. Matthews.

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

Atomic Power Station, Lincoln County, Maine

Date of amendment request: July 24, 1995

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) 3.12.B by adding an Exception to

permit a once-per-operating cycle 10 day restoration time for Remedial

Action statement 3.12.B.2. The extended restoration time would allow

maintenance to be completed on the emergency diesel generators. In

addition, the Basis of TS 3.12 is supplemented in support of the

proposed amendment.

Basis for proposed no significant hazards consideration

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

provided its analysis of the

[[Page 54722]]

issue of no significant hazards consideration. The NRC staff has

reviewed the licensee's analysis against the standards of 10 CFR

50.92(c). The NRC staff's review is presented below:

1. The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The emergency diesel generators (EDG) are not accident initiators

for any accident previously evaluated, nor does the proposed change

affect any of the assumptions used in the deterministic safety

analyses. To evaluate the effect of the proposed extended restoration

time of the EDGs fully, probabilistic safety analysis (PSA) methods

were used. The results of these analyses show no significant increase

in core damage frequency. Thus, the proposed change does not involve a

significant increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The proposed change does not alter the design, configuration, or

method of operation of the plant. Therefore, the proposed change does

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

any accident previously evaluated.

3. The proposed amendment does not involve a significant reduction

in a margin of safety.

The proposed change does not affect system or component limiting

conditions for operation, or the bases used in the deterministic

analyses to establish the margin of safety. The PSA evaluations used to

evaluate the proposed change demonstrated that the changes are either

risk neutral or risk beneficial. Thus the proposed change does not

involve a significant reduction in a margin of safety.

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

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

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Wiscasset Public Library, High

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

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

Power Company, 329 Bath Road, Brunswick, ME 04011.

NRC Project Director: Phillip F. McKee.

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

Atomic Power Station, Lincoln County, Maine

Date of amendment request: August 15, 1995.

Description of amendment request: The proposed amendment would

allow reduced power operation as a function of total reactor coolant

flow, for flow reductions as much as 5 percent below the currently

specified minimum flow. Specifically, operation would be allowed with

total flow rates below 360,000 gpm, if rated thermal power is reduced

by 1.5 percent for each 1.0 percent that total reactor coolant flow is

reduced.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The staff's review 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 amendment does not involve any changes in the

configuration of the reactor coolant system. Thus, precursors to

accidents previously evaluated are unchanged. The 5.0 percent reduction

in reactor coolant flow introduces a relatively minor change to the

overall plant heat balance, which is conservatively offset by the

proposed requirement to reduce rated thermal power by 1.5 percent for

each 1.0 percent reduction in reactor coolant system flow. Analysis by

the licensee shows that a 1.0 percent reduction in rated thermal power

for every 1.0 percent reduction in reactor coolant system flow is

sufficient to ensure that the current departure from nuclear boiling

ratio is maintained. The licensee asserts that achieving the reduced

power and other, related limits, within 24-hours of a subject flow

reduction will not significantly increase the probability or

consequences of an accident previously evaluated. Thus, the proposed

amendment does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The proposed amendment does not involve any modifications or

additions to plant equipment, and the design and operation of the plant

are not affected. The reduction in rated thermal power, reactor

protection system trip points, and operating limits conservatively

offset the reduction in reactor coolant system flow. Plant operating

conditions remain bounded by Final Safety Analysis Report (FSAR)

Chapter 14, Safety Analysis. Thus, the proposed amendment does not

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

accident previously evaluated.

3. The proposed amendment does not involve a significant reduction

in a margin of safety.

Plant rated power is conservatively reduced, consistent with the

reactor coolant flow reduction. The power reduction is specifically

designed to maintain the margin to the specified acceptable fuel design

limit on the departure from nuclear boiling ratio (DNBR), as defined in

MY TS 2.2. The licensee has evaluated this margin using the

methodologies identified in Maine Yankee Technical Specification 5.14.

The reduction in power level, operating limits, and reactor protection

system setpoints ensures that the DNBR margin is maintained for those

FSAR Chapter 14 events that rely on automatic reactor trip protection.

Power level reductions ensure that the total sensible heat in the

reactor coolant system is conservative for those events dependent on

initial system energy. Thus, the proposed change does not involve a

significant reduction in a margin of safety.

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

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

that this amendment request involves no significant hazards

consideration.

Local Public Document Room location: Wiscasset Public Library, High

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

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

Power Company, 329 Bath Road, Brunswick, ME 04011.

NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit Nos. 2, New London, Connecticut

Date of amendment request: September 19, 1995.

Description of amendment request: The proposed amendment would

reduce the frequency of the surveillance interval of the Safety

Injection Tanks (SITs) boron concentration from once per 31 days to

once per 6 months.

Basis for proposed no significant hazards consideration

determination:

[[Page 54723]]

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:

Pursuant to 10CFR50.92, Northeast Nuclear Energy Company (NNECO)

has reviewed the proposed change. NNECO concludes that the change

does not involve a significant hazards consideration since the

proposed change satisfies the criteria in 10CFR50.92(c). That is,

the proposed change does not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The revised Safety Injection Tank (SIT) surveillance

requirements meet all design and performance criteria. The change

has no [e]ffect on the ability of the SIT to perform its designed

function of providing borated water to the core following a

depressurization as a result of a Loss of Coolant Accident (LOCA).

Therefore, the changes to SIT surveillance requirements will not

increase the probability or consequences of an accident previously

evaluated.

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

from any previously analyzed.

The revised SIT surveillance requirements meet all design and

performance criteria. The change has no [e]ffect on the ability of

the SIT to perform its design function of providing borated water to

the core following a depressurization as a result of a LOCA. The

change to the SIT surveillance requirement will not create the

possibility of a new or different kind of accident from any

previously analyzed.

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

The boron concentration of the SIT will not be affected by the

change to the surveillance requirement. The boron concentration

within the SIT will continue to be monitored on a basis consistent

with the historical performance. These changes will have no impact

on the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit Nos. 2, New London, Connecticut

Date of amendment request: September 29, 1995.

Description of amendment request: The proposed amendment would

modify the Technical Specifications 3.4.2.1, 3.4.2.2, 3.7.1.1, and

Table 4.7-1.

The proposed license amendment combines three separate changes to

the Millstone Unit No. 2 Technical Specifications which pertain to

safety valves. The first proposed modification would expand the as-

found tolerance of the lift setting pressure for the pressurizer and

the main steam safety valves from the current value of plus or minus 1

percent to plus or minus 3 percent. Clarifications have also been

proposed by specifying that the lift setting pressure shall be

determined at normal operating conditions and shall be set within plus

or minus 1 percent of the required lift setting. The second portion of

the modification would eliminate the need to verify the main steam

safety valve orifice size. The third modification would modify the main

steam safety valve action statement to reflect that if a main steam

safety valve is inoperable and compensating action cannot be taken that

the plant must be brought to hot shutdown (Mode 4) in 12 hours instead

of cold shutdown (Mode 5) in 30 hours.

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 changes

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The change in the as-found pressurizer safety valve tolerance

will not increase the probability of occurrence of any of the design

basis accidents. Even with the larger tolerance, the setpoint will

provide margin to normal operation, the reactor setpoint, and PORV

[power-operated relief valve setpoint]. This minimizes the

challenges to safety valves and assures that there is no increase in

the probability of an inadvertent opening of a pressurizer safety

valve. Similarly, even with the increase in allowed as-found

tolerance for the main steam safety valves, the setpoints will still

provide margin to normal operation. Thus, there is no impact on the

probability of an inadvertent opening of a steam generator safety

valve.

The loss of load event and the inadvertent closure of one main

steam isolation valve have been reanalyzed to show that even with a

[plus or minus] 3 percent tolerance for the pressurizer safety

valves and the main steam safety valves, that both the peak RCS

[reactor coolant system] pressure and the peak steam generator

pressure remain below 110 percent of design. Thus, even with the

larger as-found tolerances, the margin of safety for RCS and steam

generator overpressurization is maintained.

The steam generator tube rupture has been reanalyzed to take

into account the [plus or minus] 3 percent as-found tolerance and to

extend the margin for operator action to one hour. A comparison of

the calculated doses shows that with the new assumptions, there

would be a very small increase in calculated doses. The increased

calculated doses, however, remain well below the Standard Review

Plan acceptance criteria.

The proposed change in the shutdown mode does not impact the

probability or consequences of an accident previously evaluated. The

proposed change makes the action required for inoperable main steam

safety valves consistent with the modes that the technical

specification is applicable and would not modify the assumptions

made in any accident previously analyzed.

The change to delete the main steam safety valve orifice size

from technical specifications has no impact on any design basis

accident analysis.

Based upon these evaluations, it is concluded that the proposed

changes do not significantly increase the probability or

consequences of any design basis accident.

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

from any previously analyzed.

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

different kind of accident from any previously analyzed.

The proposed changes do not change the as-left setpoints. The

change in as-found tolerances for the safety valves is being made to

reflect the results of past surveillances that indicate that the

setpoints can drift more than the current criteria. However, there

is no change in the plant configuration or in as-left setpoints.

The proposed change which requires the plant to go to Mode 4 in

12 hours instead of Mode 5 in 30 hours if the action statement is

not met, is consistent with the applicable modes of the technical

specification (i.e., the technical specification is not applicable

in Mode 4). No new or different kind of accident from those

previously analyzed can be postulated as a result of this proposed

change.

Thus, the changes do not create the possibility of a new or

different kind of accident from any previously analyzed.

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

As discussed above, the loss of load event and the inadvertent

closure of one main steam isolation valve have been reanalyzed to

show that even with a [plus or minus] 3 percent tolerance for the

pressurizer safety valves and the main steam safety valves, that

both the peak RCS pressure and the peak steam generator pressure

remain below 110 percent of design. Thus, even with the larger as-

found tolerances, the margin of safety for RCS and steam generator

overpressurization

[[Page 54724]]

is maintained. In addition, the steam generator tube rupture has been

reanalyzed with a [plus or minus] 3 percent tolerance on the steam

generator safety valves and the results show an insignificant

increase in the calculated doses.

The proposed change also directs the operator to bring the plant

to hot shutdown instead of cold shutdown to be consistent with the

applicable modes of the technical specification. There is no impact

on the assumptions made or the results of any accident previously

analyzed.

Therefore, it is concluded that the changes do not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

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

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

Pennsylvania

Date of amendment request: September 18, 1995

Description of amendment request: The proposed amendment would

relocate Fire Protection requirements from the Technical Specifications

to the Technical Requirements Manual. In addition, the proposed

amendment would revise Technical Specifications to include the

requirement for a program and procedure to implement the Technical

Requirements Program, and also revises Technical Specifications to add

the requirement for the Plant Operations Review Committee to review all

proposed changes to the Technical Requirements Program and to forward

copies of reviewed changes to the Susquehanna Review Committee.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change relocates the provisions of the Fire

Protection Program that are contained in the Technical

Specifications and places them in the Technical Requirements Manual.

No requirements are being added or deleted. Review and approval of

those portions of the Fire Protection Program contained in the

Technical Requirements Manual and revisions thereto will be the

responsibility of the Plant Operations Review Committee just as it

was their responsibility to review changes to the fire protection

Limiting Condition for Operation and Surveillance Requirements when

they were part of the Technical Specifications. Requiring review by

the Plant Operations Review Committee reinforces the importance of

the Technical Requirements Manual and the requirements controlled by

it and assures a multidisciplined review. Approved Technical

Requirements or changes thereto are provided to the Susquehanna

Review Committee for information. No design basis accidents are

affected by the change, nor are safety systems adversely affected by

the change. Therefore, there is no impact on the probability of

concurrence [occurrence] or the consequences of any design basis

accidents.

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

from any previously evaluated.

The proposed changes relocate the provisions of the Fire

Protection Program that are contained in the Technical

Specifications and places them in the Technical Requirements Manual.

No requirements are being added or deleted by the Technical

Requirements Manual. There are no new failure modes associated with

the proposed changes. Therefore, since the plant will continue to

operate as designed, the proposed changes will not modify the plant

response to an accident.

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

No change is being proposed for the Fire Protection Program

requirements themselves. The relevant Technical Specifications are

being relocated, and the requirements contained therein are being

incorporated into the Technical Requirements Manual. Plant

procedures will continue to provide the specific instructions

necessary for the implementation of the requirements, just as when

the requirements resided in the Technical Specifications. Fire

Protection Program changes will be subject to the provisions of 10

CFR 50.59 and the current fire protection license condition. As

such, the changes do not directly affect any protective boundaries

nor does it [do they] impact the safety limits for the boundary.

Review and approval of those portions of the Fire Protection Program

contained in the Technical Requirements Manual and the revisions

thereto will be the responsibility of the Plant Operations Review

Committee just as it was their responsibility to review changes to

the fire protection Limiting Condition for Operation and

Surveillance Requirements when they were part of the Technical

Specification. Approved Technical Requirements or changes thereto

are provided to the Susquehanna Review Committee for information.

Thus, there are no adverse impacts on the protective boundaries,

safety limits, or margin of safety.

Since operability and surveillance requirements will remain in a

controlled document, the changes do not reduce the effectiveness of

Technical Specification requirements. Any changes to the Fire

Protection Program requirements will be made in accordance with the

provisions of 10 CFR 50.59 and the fire protection license

condition.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

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

Pennsylvania 18701.

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

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

NRC Project Director: John F. Stolz.

Virginia Electric and Power Company, Docket No. 50-338, North Anna

Power Station, Unit No. 1, Louisa County, Virginia

Date of amendment request: September 19, 1995.

Description of amendment request: The proposed change would revise

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

Units 1 & 2 (NA-1 & 2). Specifically, the proposed changes would revise

TS Limiting Condition for Operation (LCO) 3.7.1.1 Action Statements, TS

Table 3.7-1, dually entitled ``Maximum Allowable Power Range Neutron

Flux High Setpoint With Inoperable Steam Line Safety Valves During 3

Loop Operation'' and ``Maximum Allowable Power Range Neutron Flux High

Setpoint With Inoperable Steam Line Safety Valves During 2 Loop

Operation,'' and the TS Bases 3/4.7.1.1, ``Safety Valves'' for NA-1 &

2. Table 3.7-1 provides the maximum allowable power range neutron flux

high setpoints with one or more main steam safety valves (MSSVs)

inoperable during two loop and three loop operation. The proposed

changes provide more conservative power range neutron flux high

setpoints calculated utilizing the Westinghouse Electric Corporation

(Westinghouse) recommended methodology and delete the information for

setpoints for two loop operation. The proposed changes also revise the

TS Bases to reflect the

[[Page 54725]]

methodology used to establish the new setpoints, and delete the LCO

Action Statement and the TS Bases for two loop operation.

Additionally, the information in Table 3.7-1 and the LCO Action

Statement associated with two loop operation have been deleted since

Virginia Electric and Power Company is prohibited by the license from

operating in this configuration.

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:

Specifically, operation of the North Anna Power Station in

accordance with the proposed Technical Specifications changes will

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

This change reduces the power level at which the reactor may be

operated with one or more main steam safety valves (MSSVs)

inoperable to ensure that the secondary system is not

overpressurized during the most severe pressurization transient of

the secondary side. There is no change to the function of the MSSVs

by the proposed change and will not alter any accident analysis

assumptions or results. The proposed changes will provide

conservative power range neutron flux high trip setpoints such that

the maximum power level allowed for operation with inoperable MSSVs

is below the heat removing capability of the operable MSSVs.

Therefore, this change will not increase the probability of an

accident.

This change is consistent with the current accident analysis

assumptions for the MSSVs and does not change the containment

response for any design basis event. Therefore, no change in the

mitigation of an accident will result from this proposed change and

no change will occur in the consequences of any accident currently

analyzed.

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

from any accident previous[ly] evaluated.

Since the implementation of the proposed changes to the

setpoints will not require hardware modifications (i.e., alterations

to plant configuration), operation of the facilities with these

proposed Technical Specifications does not create the possibility

for any new or different kind of accident which has not already been

evaluated.

The proposed revision to the Technical Specifications will not

result in any physical alteration to any plant system, nor would

there be a change in the method by which any safety-related system

performs its function. The design and operation of the main steam

system is not being changed.

These changes do not change the design, operation, or failure

modes of the main steam system. Therefore, the proposed change does

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

from any accident previously evaluated.

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

The proposed change reduces the total energy of the reactor

coolant system that will ensure the ability of the MSSVs to perform

their intended function as assumed in the current accident analyses.

Correcting this non conservatism restores the margin of safety to

what was originally envisioned. In addition, the results of the

accident analyses which are documented in the UFSAR bound operation

under the proposed changes, so that there is no safety margin

reduction. Therefore, the proposed change does not involve a

significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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., Riverfront Plaza,

East Tower, 951 E. Byrd Street, Richmond, Virginia 23219.

NRC Project Director: David B. Matthews.

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: September 19, 1995.

Description of amendment request: The proposed change would revise

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

Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed change would

increase the surveillance test interval for the turbine reheat stop and

intercept valves to once per 18 months and extend the visual and

surface inspection interval to 60 months. The proposed change would

also remove the requirement to perform additional visual and surface

inspections on the remaining turbine overspeed protection system

control valves of that type when unacceptable flaws or excessive

corrosion are identified which can be directly attributed to a service

condition specific to the inspected valve.

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:

Specifically, operation of the North Anna Power Station in

accordance with the proposed Technical Specifications changes will

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

No new or unique accident precursors are introduced by these

changes in surveillance requirements. The probability of turbine

missile ejection with an extended 18-month test interval for the

reheat stop and intercept valves has been determined to be within

the applicable acceptance criteria.

The heavy hub design of the turbine rotors provides further

assurance that the probability of the ejection of destructive

missiles remains minimal.

Based upon the results of the probabilistic evaluation, the

probability of a turbine generated missile is less than 10-5

per year which the Commission has endorsed as the acceptable level

for turbine operation.

The reheat stop and intercept valve inspection interval

extension and the elimination of the additional visual/surface

inspections do not change the design, operation, or failure modes of

the valves and other components in the turbine overspeed protection

system.

Therefore, these changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The demonstrated high reliability of the turbine reheat stop and

intercept valves and the verification of the operability of the

other turbine control valves provide adequate assurance that the

turbine overspeed protection system will operate as designed, if

needed. Turbine reheat stop and intercept valve testing performed to

date has demonstrated the reliability of these valves. In addition,

the operability of the other turbine valves (i.e., turbine throttle

valves and governor valves) will continue to be verified every 31

days or as required by the Technical Specifications.

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

from any accident previous[ly] evaluated.

Since the implementation of the proposed change to the

surveillance requirements will not require hardware modifications

(i.e., alterations to plant configuration), operation of the

facilities with these proposed Technical Specifications does not

create the possibility for any new or different kind of accident

which has not already been evaluated in the Updated Final Safety

Analysis Report (UFSAR). In addition, the results of the

probabilistic evaluation indicate that no additional transients have

been introduced.

The proposed revision to the Technical Specifications will not

result in any physical alteration to any plant system, nor would

there be a change in the method by which any safety-related system

performs its function. The design and operation of the turbine

overspeed protection and turbine control systems are not being

changed.

The proposed Technical Specifications changes do not affect the

design, operation, or failure modes of the valves and other

components of the turbine overspeed protection system.

[[Page 54726]]

Therefore, the proposed change does not create the possibility

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

evaluated.

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

The proposed changes would not reduce the margin of safety as

defined in the basis for any Technical Specifications. The design

and operation of the turbine overspeed protection and turbine

control systems are not being changed and the operability of the

turbine reheat stop and intercept valves will be demonstrated on a

refueling outage basis. In addition, the results of the accident

analyses which are documented in the UFSAR continue to bound

operation under the proposed changes, so that there is no safety

margin reduction. Therefore, the proposed change does not involve a

significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: David B. Matthews.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of amendment request: August 15, 1995.

Description of amendment request: This application to revise the

Braidwood, Unit 1, Technical Specifications (TSs) proposes to continue

to use the voltage-based repair criteria which were added to the

Braidwood, Unit 1, TSs by a license amendment issued on August 18,

1994. This August 15, 1995, request will be considered by the staff

only in the event that the staff can not reach a timely decision on

your pending request for license amendments dated September 1, 1995, to

raise the present lower voltage repair limit from 1.0 volt to 3.0

volts.

Date of publication of individual notice in Federal Register:

October 5, 1995 (60 FR 52222).

Expiration date of individual notice: November 6, 1995.

Local Public Document Room location: 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 amendment request: September 15, 1995.

Description of amendment request: To close out open items

identified by the NRC staff's review of the upgrade of sections 1.0, 3/

4.4, 3/4.10, and 5.0 of the Dresden and Quad Cities Technical

Specifications to the BWR Standard Technical Specifications.

Date of publication of individual notice in Federal Register:

October 5, 1995 (60 FR 52220).

Expiration date of individual notice: November 6, 1995.

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. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: September 20, 1995.

Description of amendment request: The proposed amendment would

upgrade the Quad Cities TS to the Standard Technical Specifications

(STS) contained in NUREG-0123. The Technical Specification Upgrade

Program (TSUP) is not a complete adaption of the STS. The TS upgrade

focuses on (1) integrating additional information such as equipment

operability requirements during shutdown conditions, (2) clarifying

requirements such as limiting conditions for operation and action

statements utilizing STS terminology, (3) deleting superseded

requirements and modifications to the TS based on the licensee's

responses to Generic Letters (GL), and (4) relocating specific items to

more appropriate TS locations. The September 20, 1995, application

proposed to upgrade only Section 6.0 (Administrative Controls) of the

Quad Cities TS.

Date of publication of individual notice in Federal Register:

October 5, 1995 (60 FR 52226).

Expiration date of individual notice: November 6, 1995.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

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

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

Date of amendment request: September 5, 1995.

Description of amendment request: The proposed amendment would

modify the Appendix A Technical Specifications (TSs) for the Turbine

Cycle Safety Valves. Specifically, the proposed amendment would change

Seabrook Station Appendix A Technical Specification Table 3.7-1 to

reduce the maximum allowable Power Range Neutron Flux--High setpoints

with inoperable Main Steam Safety Valves (MSSVs) and Table 3.7-2 to

reduce the opening setpoints of the MSSVs.

Date of publication of individual notice in Federal Register:

October 2, 1995 (60 FR 51505).

Expiration date of individual notice: November 1, 1995.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

[[Page 54727]]

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: August 3, 1995.

Brief description of amendments: These amendments add the

analytical method supplement entitled ``Fuel Rod Maximum Allowable Gas

Pressure,'' CEN-372-P-A, dated May 1990, and its associated NRC Safety

Evaluation, dated April 10, 1990, to the list of analytical methods in

Technical Specification 6.9.1.10 used to determine the Palo Verde

Nuclear Generating Station core operating limits.

Date of issuance: October 4, 1995.

Effective date: October 4, 1995, to be implemented prior to startup

from RF06 for Units 1 and 2, and RF5 for Unit 3.

Amendment Nos.: Unit 1--Amendment No. 101; Unit 2--Amendment No.

89; Unit 3--Amendment No. 72.

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

amendments revised the Technical Specifications.

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

45173) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 4, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

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

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

Maryland

Date of application for amendments: June 2, 1995.

Brief description of amendments: The amendments revise the

tolerances for the pressurizer safety valve as-found acceptance

criterion.

Date of issuance: September 26, 1995.

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

30 days.

Amendment Nos.: 206 and 184.

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

revised the Technical Specifications.

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

35060) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated September 26, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

Maryland

Date of application for amendments: January 31, 1995.

Brief description of amendments: The amendments revise the

Technical Specifications (TSs) to increase the amount of Trisodium

Phosphate Dodecahydrate located in the containment sump baskets which

is required to be verified by TS surveillance. The test requirements

for verifying that the appropriate pH (acidity/alkalinity) would be

maintained in the containment sump water following a design-basis

accident are moved from the TSs to the TS Bases section; however, the

requirement to perform the test remains in the TSs. The associated TS

Bases sections are updated to reflect the changes.

Date of issuance: October 5, 1995.

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

30 days.

Amendment Nos.: 207 and 185.

Facility Operating License No. DPR-53 and DPR-69: Amendment revised

the Technical Specifications.

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

14016) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated October 5, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678

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

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

Date of application for amendments: March 8, 1995, as supplemented

on June 1, 1995.

Brief description of amendments: The amendments revise the

secondary undervoltage setpoint.

Date of issuance: October 2, 1995.

Effective date: October 2, 1995

Amendment Nos.: 169 and 156.

Facility Operating License Nos. DPR-39 and DPR-48: The amendments

revised the Technical Specifications.

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

45178) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 2, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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: February 18, 1994, as

supplemented June 3, November 1, December 2, December 14, and December

16, 1994, and August 25, 1995.

Brief description of amendment: The amendment revises the

surveillance intervals for the Boric Acid Tank Level, the Service Water

Inlet Temperature Monitor Instrument, the Boric Acid Makeup Flow

System, the Plant Noble Gas Activity Monitor, the Condenser Evacuation

System Activity Monitor, the Low Turbine Auto Stop Oil Pressure Trip,

the 6.9 kv Undervoltage Monitor, the Sampler Flow Rate Monitor, and the

Refueling Water Storage Tank.

Date of issuance: October 12, 1995.

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

30 days.

Amendment No.: 184.

[[Page 54728]]

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

Technical Specifications.

Date of initial notice in Federal Register: April 28, 1994 (59 FR

22003) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 12, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment: March 17, 1995.

Brief description of amendment: The amendment revises requirements

associated with channel functional tests of the core protection

calculator following a high temperature alarm.

Date of issuance: October 11, 1995.

Effective date: October 11, 1995, to be implemented within 30 days.

Amendment No.: 168.

Facility Operating License No. NPF-6. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 2, 1995 (60 FR

39437) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 11, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 18, 1991, as supplemented by

letters dated March 16, and December 2, 1994, and March 9, and August

30, 1995.

Brief description of amendment: The amendment changes the Appendix

A TSs by subdividing TS 3/4.7.6, ``Control Room Air Conditioning

System,'' into five separate TSs covering the following three distinct

functions: control room emergency air filtration, control room air

temperature, and control room isolation and pressurization. The

amendment also changes the Bases sections of the TS to reflect the

above changes.

Date of issuance: October 4, 1995.

Effective date: October 4, 1995.

Amendment No.: 115.

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

Technical Specifications.

Date of initial notice in Federal Register: September 4, 1991 (56

FR 43808) and July 6, 1995 (60 FR 29875).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 4, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: June 1, 1995, as supplemented

August 23, 1995.

Brief description of amendment: The amendment changes the Technical

Specifications to relocate the procedural details of the Radiological

Effluent Technical Specifications to the Offsite Dose Calculation

Manual. With these changes, the specifications related to RETS

reporting requirements were simplified and changes to the definition of

the ODCM were made to make the definition consistent with the

amendment.

Date of Issuance: October 2, 1995.

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

120 days.

Amendment No.: 197.

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

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

35078) The August 23, 1995, letter provided supplemental information

that did not change the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated October 2, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (Regional Depository) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: January 16, 1995, as

supplemented June 22 and September 20, 1995.

Brief description of amendment: The amendment revises the Technical

Specifications for TMI-1 to incorporate seven improvements from the

Revised Standard Technical Specifications for Babcock & Wilcox Nuclear

Power Plants (NUREG-1430). The amendment also changes the Bases

incorporating the results of analyses to support allowance for drift of

the Pressurizer Code Safety Valve setpoint. The remaining portion of

the request relating to revisions to Control Room Emergency Ventilation

system are being reviewed separately.

Date of Issuance: October 10, 1995.

Effective date: October 10, 1995.

Amendment No.: 198.

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

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

14021). The June 22 and September 20, 1995, letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated October 10, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (Regional Depository) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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 application for amendments: November 12, 1993, as

supplemented November 18, 1994, May 30, 1995, and August 8, 1995.

Brief description of amendments: The amendments delete from the

Technical Specifications the sections and tables entitled ``Component

Cyclic or Transient Limits'' and relocate the information to the

Updated Final Safety Analysis Report.

Date of issuance: September 28, 1995.

Effective date: September 28, 1995, with full implementation within

45 days.

Amendment Nos.: 201 and 186.

Facility Operating License Nos. DPR-58 and DPR-74. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 22, 1993 (58

FR 67849). The November 18, 1994, May 30, 1995, and August 8, 1995,

supplements provided clarifying information and corrections to

additional pages which referenced the table to be deleted. This

information was within the scope of the original

[[Page 54729]]

application and did not change the staff's initial proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated September 28, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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 application for amendments: May 26, 1995.

Brief description of amendments: The amendments modify Technical

Specification Sections 3/4.3.1 and 3/4.3.2 and their accompanying

Bases, to relocate the tables of response time limits for the reactor

trip system and engineered safety feature acutation system

instrumentation to the Updated Final Safety Analysis Report.

Date of issuance: October 10, 1995.

Effective date: October 10, 1995.

Amendment Nos.: 202 and 187.

Facility Operating License Nos. DPR-58 and DPR-74. Amendments

revised the Technical Specifications.

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

35082) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 10, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

Atomic Power Station, Lincoln County, Maine

Date of application for amendment: May 5, 1995.

Brief description of amendment: The amendment revises the

surveillance frequency of radiation area, and effluent and process

monitors from monthly to quarterly; and the required frequency for

minimum exercise of control element assemblies also from monthly to

quarterly.

Date of issuance: October 2, 1995.

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

within 30 days.

Amendment No.: 153.

Facility Operating License No. DPR-36: Amendment revised the

Technical Specifications and/or License.

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

45179). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 2, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wiscasset Public Library, High

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

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of application for amendment: June 8, 1994, as superseded by

letter dated April 20, 1995, and supplemented by letter dated August

18, 1995.

Brief description of amendment: The amendment revises Sections 3.7/

4.7, which pertain to the standby gas treatment system (SGTS) and

secondary containment. The amendment revises the surveillance

requirements for both SGTS and the secondary containment and revises

the performance requirements for the SGTS filters and process stream

electric heaters.

Date of issuance: October 2, 1995.

Effective date: October 2, 1995.

Amendment No.: 94.

Facility Operating License No. DPR-22. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 20, 1994 (59 FR

37075). The April 20 and August 18, 1995, submittals provided

clarifying information within the scope of the original submittal and

did not change the staff's initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 2, 1995.

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 Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,

Minnesota.

Date of application for amendments: July 11, 1994, as supplemented

April 18, 1995 (supersedes the February 10, 1993, application).

Brief description of amendments: The amendments change license

condition 2.C.(4) of each license to conform to the standard fire

protection license condition as stated in Generic Letter 86-10. In

addition, the amendments delete fire protection program elements from

the Technical Specifications and incorporate, by reference, the NRC-

approved Fire Protection Program and major commitments, including the

fire hazards analysis, into the Updated Safety Analysis Report.

Date of issuance: October 6, 1995.

Effective date: October 6, 1995, with full implementation within 30

days.

Amendment Nos.: 120 and 113.

Facility Operating License Nos. DPR-42 and DPR-60. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 21, 1994 (59

FR 65818). The April 18, 1995, letter provided clarifying information

within the scope of the original submittal and did not change the

staff's initial proposed no significant hazards consideration

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

contained in a Safety Evaluation dated October 6, 1995.

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.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco

Nuclear Generating Station, Sacramento County, California

Date of application for amendment: June 20, 1995.

Brief description of amendment: This amendment modifies the

technical specifications on spent fuel storage building load handling

limits to allow the placement of the top shield plug on a dry shielded

canister containing spent fuel which is being prepared for transfer to

the Rancho Seco Independent Spent Fuel Storage Installation.

Date of issuance: October 5, 1995.

Effective date: October 5, 1995.

Amendment No.: 123.

Facility Operating License No. NPF-1: The amendment revised the

Technical Specifications.

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

45184). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 5, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Central Library, Government

Documents, 828 I Street, Sacramento, California 95814.

[[Page 54730]]

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: June 30, 1995, as supplemented

on August 11, 1995.

Brief description of amendment: The amendment revises the Technical

Specifications (TS) for the pressurizer power operated relief valves to

follow the NRC's guidance of Generic Letter 90-06 (Generic Issue 70),

and the improved Westinghouse Standard TS (NUREG-1431, Rev. 1).

Date of issuance: September 18, 1995.

Effective date: September 18, 1995.

Amendment No.: 129.

Facility Operating License No. NPF-12. Amendment revises the TS.

Date of initial notice in Federal Register: Auust 16, 1995 (60 FR

42608).

The August 11, 1995, supplemental letter corrected an error in the

original submittal and did not change the initial proposed no

significant hazards consideration. The Commission's related evaluation

of the amendment is contained in a Safety Evaluation dated September

18, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

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

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

Alabama

Date of amendments request: August 17, 1994, as supplemented by

letters dated June 15 and August 11, 1995.

Brief Description of amendments: The amendments eliminate periodic

pressure sensor response time testing surveillance requirements for

specific Reactor Trip System and Engineered Safety Feature Actuation

System instrumentation specified in Technical Specification Sections

4.3.1.3 and 4.3.2.3.

Date of issuance: September 28, 1995.

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

30 days from the date of issuance.

Amendment Nos.: 116 and 108.

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

the Technical Specifications.

Date of initial notice in Federal Register: September 28, 1994 (59

FR 49434) The June 15 and August 11, 1995, letters provided clarifying

information that did not change the scope of the August 17, 1994,

application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 28, 1995.

No significant hazards consideration comments received: No.

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

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

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M.

Farley Nuclear Plant, Unit 1, Houston County, Alabama

Date of amendment request: December 7, 1994, as supplemented by

letter dated May 31, 1995.

Brief Description of amendment: The amendment revised Farley Unit 1

Technical Specifications 4.4.6.2, 4.4.6.4, 4.4.6.5, 3.4.7.2, and 3.4.9

for Cycle 14 operation to permit the use of steam generator tube repair

criteria for defects confined within the thickness of the tube support

plate.

Date of issuance: September 28, 1995.

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

to the start of Unit 1, Cycle 14 operation.

Amendment No.: 117.

Facility Operating License No. NPF-2: Amendment revised the

Technical Specifications.

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

FR 8754) The May 31, 1995, letter provided clarifying information that

did not change the scope of the December 7, 1994, application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 28, 1995.

No significant hazards consideration comments received: No.

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

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

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: August 7, 1995 (TS 95-18).

Brief description of amendments: The amendments revise the titles

of various administrative positions found in Section 6.0 of the

Technical Specifications.

Date of issuance: October 2, 1995.

Effective date: October 2, 1995,

Amendment Nos.: 212 and 202.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the technical specifications.

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

45186) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 2, 1995.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: August 7, 1995 (TS 95-12).

Brief description of amendments: The amendments correct various

editorial errors in the text of the technical specifications and remove

provisions that have expired or are no longer applicable.

Date of issuance: October 4, 1995.

Effective date: October 4, 1995.

Amendment Nos.: 213 and 203.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the technical specifications.

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

45185) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 4, 1995.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,

Unit 1, Hamilton County, Tennessee

Date of application for amendment: July 19, 1995, superseded

September 7, 1995 and supplemented September 15 and 26, 1995 (TS 95-

15).

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 similar to guidance given in Generic Letter (GL) 95-

05, ``Voltage-Based Repair Criteria for Westinghouse Steam Generator

Tubes Affected by Outside Diameter Stress Corrosion Cracking.''

Date of issuance: October 11, 1995.

Effective date: October 11, 1995.

Amendment No.: 214.

[[Page 54731]]

Facility Operating License Nos. DPR-77: Amendment revises the

technical specifications.

Date of initial notice in Federal Register: August 1, 1995 (60 FR

39189) The letters dated September 7, 15 and 26, 1995 provided

information that did not change the initial proposed no significant

hazards consideration. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated October 11, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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: March 30, 1995, as supplemented

August 24, 1995

Brief description of amendments: The amendments revise the North

Anna 1 and 2 Technical Specifications to allow one of the two service

water loops to be isolated from the component cooling water head

exchangers during power operations in order to refurbish the isolated

service water headers.

Date of issuance: October 11, 1995.

Effective date: October 11, 1995.

Amendment Nos.: 194 and 175.

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

the Technical Specifications.

Date of initial notice in Federal Register: May 10, 1995 (60 FR

24923). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 11, 1995.

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.

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of application for amendment: December 6, 1993.

Brief description of amendment: The amendment changes the

surveillance requirements in Technical Specification 4.6.6.1.b.3 to

provide more appropriate acceptance criteria for demonstrating

operability of the primary containment hydrogen recombiner systems.

Date of issuance: October 5, 1995.

Effective date: October 5, 1995, to be implemented within 30 days

of issuance.

Amendment No.: 142.

Facility Operating License No. NPF-21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 6, 1994 (59 FR

34670). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 5, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

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

Generating Station, Coffey County, Kansas

Date of amendment request: May 24, 1994, as supplemented by letter

dated April 6, 1995.

Brief description of amendment: This amendment revises the

technical specifications (TS) to implement the NRC's revised 10 CFR

50.36 on technical specification improvements for nuclear power

reactors. Specifications that do not meet any of the four criteria or

regulatory requirements related to inclusion in the TS are relocated to

Chapter 16 of the Updated Safety Analysis Report.

Date of issuance: October 2, 1995.

Effective date: October 2, 1995, to be implemented within 120 days

from the date of issuance.

Amendment No.: 89.

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

Technical Specifications.

Date of initial notice in Federal Register: July 6, 1994 (59 FR

34671). The April 6, 1995, supplemental letter provided additional

clarifying information and did not change the initial no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 2, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

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

66621.

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

Generating Station, Coffey County, Kansas

Date of amendment request: July 25, 1995.

Brief description of amendment: The amendment deletes a clause from

Section 4.0.5a, ``Surveillance Requirements for Inservice Inspection

and Testing Program.'' This clause required prior NRC approval before

implementation of a relief request upon finding an ASME Code

requirement impractical because of prohibitive dose rates or

limitations in the design, construction, or system configuration.

Date of issuance: October 4, 1995.

Effective date: October 4, 1995, to be implemented within 30 days

of issuance.

Amendment No.: 90.

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

Technical Specifications.

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

45191). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 4, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

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

66621

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

[[Page 54732]]

media to provide notice to the public in the area surrounding a

licensee's facility of the licensee's application and of the

Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

application for amendment, (2) the amendment to Facility Operating

License, and (3) the Commission's related letter, Safety Evaluation

and/or Environmental Assessment, as indicated. All of these items are

available for public inspection at the Commission's Public Document

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

the local public document room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By November 24, 1995, 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

[[Page 54733]]

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 the

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

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

Michigan

Date of application for amendment: October 2, 1995.

Brief description of amendment: The amendment revises the Technical

Specifications to allow deferral until the next plant outage of certain

portions of logic system functional surveillance testing for the diesel

generator 480-volt load sequencer and output breaker reclosure logic

circuitry.

Date of issuance: October 13, 1995.

Effective date: October 13, 1995, with full implementation within

45 days.

Amendment No.: 105.

Facility Operating License No. NPF-43: Amendment revises 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

October 13, 1995.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

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

2000 Second Avenue, Detroit, Michigan 48226.

NRC Project Director: Brian E. Holian, 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 application for amendments: September 30, 1995.

Brief description of amendments: The amendments increase the

setpoint tolerance of the main steam safety valves (MSSVs) from plus or

minus 1 percent to plus or minus 3 percent, with the exception that the

lowest set MSSVs would have a tolerance of -2 percent/+3 percent.

Date of issuance: October 1, 1995.

Effective date: October 1, 1995.

Amendment Nos.: Unit 1--Amendment No. 108; Unit 2--Amendment No.

107.

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

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

October 1, 1995.

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.

Dated at Rockville, Maryland, this 18th day of October 1995.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects--III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 95-26275 Filed 10-24-95; 8:45 am]

BILLING CODE 7590-01-P