Biweekly Notice

[[Page 10392]]

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 February 16, 1996, through March 1, 1996.

The last biweekly notice was published on February 28, 1996 (61 FR

7542).

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 April 12, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one

[[Page 10393]]

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.

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 16, 1994, as supplemented on

January 31, 1996.

Description of amendment request: The proposed amendment would

revise the technical specifications to eliminate periodic response time

testing requirements for selected pressure and differential pressure

sensors in the reactor trip system and engineered safety features

actuation instrumentation channels.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This change to the Technical Specifications does not result in a

condition where the design, material, and construction standards

that were applicable prior to the change are altered. The same RTS

and ESFAS instrumentation is being used; the time response

allocations/modeling assumptions in the Updated Final Safety

Analysis Report (UFSAR), Chapter 15, Accident Analyses, are still

the same; only the method of verifying time response is changed. The

proposed change will not modify any system interface and could not

increase the likelihood of an accident since these events are

independent of this change. The proposed activity will not change,

degrade or prevent actions or alter any assumptions previously made

in evaluating the radiological consequences of an accident described

in the UFSAR. Therefore, the proposed amendment does not result in

any increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This change does not alter the performance of the identified

pressure and differential pressure transmitters and switches used in

the plant protection systems. All sensors will still have response

time verified by test before placing the sensor in operational

service, and after any maintenance that could affect response time.

Changing the method of periodically verifying instrument response

for these sensors (assuring equipment operability) from time

response testing to calibration and channel checks does not result

in any design, installation, or operational changes and thus will

not create any new accident initiators or scenarios. Periodic

surveillance of these instruments will detect significant

degradation in the sensor response characteristics. Implementation

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

or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

This change does not affect the total system response time

assumed in the safety analyses. The periodic system response time

verification method for the identified pressure and differential

pressure sensors and switches is modified to allow use of (1)

historical records based on acceptable response time tests

(hydraulic, noise, or power interrupt tests), (2) inplace, onsite or

offsite (e.g. vendor) test measurements, or (3) using vendor

engineering specifications.

The method of verification still provides assurance that the

total system response is within that defined in the safety analyses,

since calibration tests will detect any degradation which might

significantly affect sensor response time. Based on the above, it is

concluded that the proposed license amendment request does not

result in a reduction in margin with respect to plant safety.

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

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

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

requested amendments involve no significant hazards consideration.

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

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

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

Wilmington, Illinois 60481.

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

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

Power Station, Unit No. 1, Shippingport, Pennsylvania

Date of amendment request: February 12, 1996

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 4.6.2.2.d to delete the reference

to the specific test acceptance criteria for the Containment

Recirculation Spray Pumps and replace the specific test acceptance

criteria with

[[Page 10394]]

reference to the requirements of the Inservice Testing (IST) Program.

In addition, the 18-month test frequency would be replaced with the

test frequency requirements specified in the IST Program. The proposed

amendment would make this TS the same as Beaver Valley Power Station,

Unit No. 2 TS 4.6.2.2.d which was revised by License Amendment No. 68

on May 3, 1995.

The proposed amendment would also revise the Bases of TS 4.6.2.2.d

for both Unit Nos. 1 and 2 to describe the proposed revision to TS

4.6.2.2.d.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The change does not result in a modification to plant equipment

nor does if affect the manner in which the plant is operated. The

Recirculation Spray System (RSS) pumps are normally in a standby

condition and only operate during accident mitigation. Since the

physical plant equipment and operating practices are not changed, as

noted above, there is no change in the probability of an accident

previously evaluated.

The proposed change, for Beaver Valley Power Station (BVPS) Unit

No. 1 only, will not lower the pump performance operability criteria

for the RSS pumps. The required values for developed pump head and

flow will continue to satisfy accident mitigation requirements and

will be maintained and controlled in the BVPS Unit No. 1 Inservice

Testing (IST) Program.

Since the proposed change does not lower the RSS pump

performance acceptance criteria, the containment depressurization

system will continue to meet its design basis requirements. The

proposed change will not impose additional challenges to the

containment structure in terms of peak pressure. The calculated

offsite does consequences of a design basis accident (DBA) will

remain unchanged since the one hour release duration remains

unchanged. Future changes to the RSS pump head and flow requirements

will be made under the 10 CFR 50.59 process to ensure that the

containment performance requirements continue to be met.

The proposed change in the RSS pump surveillance interval from

18 months to every refueling, will not affect the ability of the

pumps to perform as assumed in the Safety Analyses. The proposed

change to the Bases section, for BVPS Unit Nos. 1 and 2, will ensure

that safety analyses assumptions for assumed pump performance

continue to be met. The words ``required developed head'' will be

clearly defined to reflect that they refer to the value assumed in

the safety analysis for the recirculation spray pump's developed

head at a specific point. The proposed changes to the Index pages

are administrative in nature and do not affect plant safety.

Therefore, the proposed change does not involve a significant

increase in the consequences of an accident previously evaluated.

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

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not alter the method of operating the

plant. The recirculation spray system is an accident mitigation

system and is normally in standby. System operation would be

initiated following a containment pressure increase resulting from a

DBA. The RSS pumps will continue to provide sufficient flow to

mitigate the consequences of a DBA. RSS operation continues to

fulfill the safety function for which it was designed and no changes

to plant equipment will occur. As a result, an accident which is new

or different than any already evaluated in the Updated Final Safety

Analysis Report will not be created due to this change.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

of safety?

The surveillance requirements for demonstrating that the RSS

pumps are operable will continue to assure the ability of the system

to satisfy its design function. Therefore, the proposed change will

not affect the ability of the RSS to perform its safety function.

The containment spray system design requirement to restore the

containment to subatmospheric condition within one hour will

continue to be satisfied. This proposed change does not have any

affect on the containment peak pressure since the containment peak

pressure occurs prior to the initiation of any of the two

containment spray systems. There is no resultant change in dose

consequences since the containment will continue to reach a

subatmospheric pressure within the first hour following a DBA.

The RSS pumps' performance requirements will continue to be

controlled in a manner to ensure safety analysis assumptions are

met.

Therefore, based on the above discussion, it can be concluded

that 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 10 CFR 50.92(c) are

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001

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

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

NRC Project Director: John F. Stolz

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

Linn County, Iowa

Date of amendment request: November 30, 1995

Description of amendment request: The proposed amendment would

implement the Option I-D long-term stability solution and remove the

existing SIL-380 Rev. 1-based specifications. In addition, the proposed

change would require a plant scram be initiated should the plant enter

natural circulation conditions and would prohibit restarting a

recirculation pump while in natural circulation. The proposed change

would define natural circulation. Finally, this change would delete

Technical Specification (TS) actions and surveillance requirements

related to core plate differential pressure noise while in single

recirculation pump operation (SLO).

Basis for proposed no significant hazards consideration

determination:

As required by 10 50.91(a), the licensee has provided its analysis

of the issue of no significant hazards consideration, which is

presented below:

1) The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The implementation of the [boiling water

reactor] BWR Owner's Group long term solution Option I-D does not

modify the assumptions in the existing accident analysis. The use of

an exclusion region and the operator actions required to avoid and

minimize operation inside the region do not increase the possibility

of an accident. Licensing Topical Report, 'Evaluation of the

``Regional Exclusion with Flow-Biased APRM [average power range

monitor] Neutron Flux Scram'' Stability Solution', GENE-A000-04021-

01 (attachment 1) demonstrates that the APRM flow-biased scram

function provides a high degree of assurance that the fuel safety

limit will not be exceeded should power oscillations occur during

plant operation within the restricted region. Regional mode core

oscillations are not predicted to occur at the [Duane Arnold Energy

Center] DAEC because of its small core size and tight core inlet

orifices. Conditions for operation outside of the exclusion region

are within the assumptions of the existing accident analysis. The

operator action requirement to exit the exclusion region upon entry

minimizes the probability of an instability event occurring.

Inserting control rods or increasing recirculation flow, the

evolutions to be used to exit the region, are normal plant

maneuvers.

The proposed clarifications to explicitly direct the operator to

initiate a reactor scram

[[Page 10395]]

in the event of operation in natural circulation are conservative

and consistent with current plant operating practices. Likewise, the

proposed prohibition from starting a recirculation pump as a means

of exiting the natural circulation mode of operation is also

conservative. Therefore, the proposed license amendment does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The core plate differential pressure noise surveillances that

are performed while in single recirculation pump operation were

included in TS Amendment 119 due to NRC concerns at the

time that high core plate noise observed during [single-loop

operation] SLO at Brown's Ferry in 1985 could be an indication of

thermal hydraulic instability. [General Electric] GE has since

determined that core plate differential pressure noise is not a

cause of thermal hydraulic instability and that the noise does not

pose a safety concern. Therefore, the proposed license amendment

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2) The proposed license amendment does not create the

possibility of a new or different kind of accident from any

previously evaluated. As stated above, the proposed changes either

mandate operation within the envelope of previously analyzed plant

operating conditions or direct the operator to immediately return

the plant to within these analyzed conditions using normal plant

maneuvers. In addition, analysis has demonstrated that the APRM

flow-biased scram function provides a high degree of assurance that

the fuel safety limit will not be exceeded should power oscillations

occur during plant operation within the restricted region.

Therefore, the potential for a new or different type of accident

from those previously evaluated is not created.

The proposed clarifications to explicitly direct the operator to

initiate a reactor scram in the event of operation in natural

circulation are conservative and consistent with current plant

operating practices. Likewise, the proposed prohibition from

starting a recirculation pump as a means of exiting the natural

circulation mode of operation is also conservative. Therefore, the

potential for a new or different type of accident from those

previously evaluated is not created.

The core plate differential pressure noise surveillances that

are performed while in single recirculation pump operation were

included in TS Amendment 119 due to NRC concerns at the

time that high core plate noise observed during SLO at Brown's Ferry

in 1985 could be an indication of thermal hydraulic instability. GE

has since determined that core plate differential pressure noise is

not a cause of thermal hydraulic instability and that the noise does

not pose a safety concern. Therefore, the potential for a new or

different type of accident from those previously evaluated is not

created.

3) The proposed amendment will not reduce the margin of safety.

The combination of the proposed requirements to avoid possible

unstable conditions and the automatic flow biased high reactor flux

scram provide defense in depth to provide fuel protection. Therefore

the individual or combination of means to detect and suppress

thermal hydraulic instability supplements the margin of safety.

The proposed specification related to initiating a reactor scram

while in natural circulation is conservative. Likewise, the proposed

prohibition from starting a recirculation pump as a means of exiting

the natural circulation mode of operation is also conservative and

therefore does not constitute a reduction in the margin of safety.

The core plate differential pressure noise surveillances that

are performed while in single recirculation pump operation were

included in TS Amendment 119 due to NRC concerns at the

time that high core plate noise observed during SLO at Brown's Ferry

in 1985 could be an indication of thermal hydraulic instability. GE

has since determined that core plate differential pressure noise is

not a cause of thermal hydraulic instability and that the noise does

not pose a safety concern. Therefore, the elimination of these

surveillance tests does not constitute a reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

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

Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan,

Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869

NRC Project Director: Gail H. Marcus

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: November 16, 1995

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) to add a Limiting Condition

for Operation and surveillance test for safety related inverters and

deletes requirements for non-safety related instrument buses.

Basis for proposed no significant hazards consideration

determination:

As required by 10 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 changes will delete requirements from the Technical

Specifications (TS) for non-safety related 120 Volt a-c instrument

panels AI-42A and AI-42B, and incorporate new requirements for the

safety-related 125 Volt d-c to 120 Volt a-c inverters (A, B, C, and

D) similar to the Standard Technical Specification for Combustion

Engineering plants as contained in NUREG-1432.

TS 2.7 requires that 120 Volt instrument panels AI-42A and AI-

42B be operable whenever the reactor coolant temperature is above

300 F. Either of these instrument panels may be inoperable for up

to 8 hours or a plant shutdown is required. These instrument panels

are non-safety related and do not receive or actuate any Engineered

Safeguards Features (ESF) or Reactor Protection System (RPS) and the

panels are not required for, nor do they indicate the status of,

containment integrity. The FCS plant specific Probabilistic Risk

Assessment (PRA) model was reviewed to determine the effect of

unavailability of these instrument panels on the core damage

frequency. The results of the review show that the unavailability of

these panels is not a contributor to risk. Therefore these

instrument panels do not meet any of the four criteria contained in

10 CFR 50.36 for inclusion into TS. The operation of these panels

are controlled by plant procedures that are governed by 10 CFR

50.59.

Therefore, deletion of the requirements for AI-42A and AI-42B

from the TS would not significantly increase the probability or

consequences of an accident previously evaluated.

It is also proposed to incorporate new requirements for the

safety-related 125 Volt d-c to 120 Volt a-c inverters (A, B, C, and

D). Currently, there are no TS requirements for inoperability of the

safety-related inverters. However, if an inverter is inoperable and

its associated 120 Volt a-c instrument bus is powered by its safety-

related bypass transformer, the a-c instrument bus is considered

inoperable and an 8 hour Limiting Condition for Operation is

applied. The bus is declared inoperable even though it is being

powered from a safety related power source because this source is

not an uninterruptible power supply. Operating experience has shown

that, in many instances, 8 hours is insufficient time to

troubleshoot and conduct repairs on an inverter. FCS initiated a TS

required plant shutdown in November 1994, and again in January 1995,

due to inoperable inverters that could not be repaired in the 8

hours allowed by TS. If FCS had 24 hours to conduct repairs, a power

reduction, and the potential to challenge plant systems, would not

have been necessary.

The proposed change does not increase the probability of an

accident since loss of power to a vital bus is not an initiator of

any analyzed accident. The proposed change does not increase the

consequences of any accident since the TS currently allow one 120 V

instrument bus to be inoperable and de-energized. The proposed

change would only allow one 120 V instrument bus to be energized

from a safety related bypass source. The proposed changes do not

reduce the number of RPS or ESF actuation channels that are required

to be operable. Should a

[[Page 10396]]

loss of offsite power event occur, power to the instrument bus would

only be interrupted during the time required for the emergency

diesel generator to start and load.

The FCS plant specific PRA model was reviewed to determine the

effect of unavailability of the 120 V instrument panels supplied by

inserters A, B, C, and D on the core damage frequency. The results

of the review show that the loss of one of the panels has an

insignificant effect on the PRA model. Therefore, the proposed

change of allowing a 24 hour period with one instrument panel

powered from a interruptible power supply has a insignificant effect

on the PRA results.

Therefore, the proposed change to include specific operability

requirements for safety related inverters does not significantly

increase the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

There will be no physical alterations to the plant

configuration, changes to setpoint values, or changes to the

implementation of setpoints or limits as a result of these proposed

changes. The proposed changes do not reduce the number of RPS or ESF

actuation channels that are required to be operable. Therefore, the

proposed changes do 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 delete TS requirements for nonsafety

related instrument panels and incorporate additional operability

requirements for safety related inverters. The proposed changes do

not revise any setpoints or limits monitored by the instrument

panels or buses. In addition, a review of the FCS plant specific PRA

shows that these proposed changes are insignificant to core damage

frequency. Therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502

NRC Project Director: William H. Bateman

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: February 1, 1996

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to allow an increase in the

initial nominal enrichment limit of fuel assemblies to be stored in the

spent fuel pool.

Basis for proposed no significant hazards consideration

determination:

As required by 10 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 change to the Technical Specifications to increase

the enrichment limit for fuel assembly storage requirements does not

involve a significant increase in the probability of an accident.

The enrichment limit is not a precursor to any analyzed event and

therefore cannot impact probability.

The safety evaluation for the existing Spent Fuel Pool (SFP)

storage racks was approved by the NRC in Amendment 155 (TAC M85116).

This amendment approved the current limit on fuel enrichment, and

the mechanical, structural, and thermal/hydraulic design of the fuel

racks. This amendment also evaluated the radiological consequences

of a fuel handling accident with fuel enrichments equivalent to the

proposed change. The proposed change will not impact this previously

approved evaluation with the exception of the nuclear criticality

analysis. The nuclear criticality analysis supporting the proposed

change used calculational methods conforming to NRC guidance,

industry codes, standards, and specifications. In meeting the

acceptance criteria for criticality in the SFP, such that keff

is always less than or equal to O.95 at a 95%/95% probability

tolerance level, the proposed change from 4.2 weight percent (w/o)

to 4.5 w/o Uranium-235 (U235) does not involve an increase in

the consequences of an accident previously evaluated.

Therefore, it is concluded that the proposed change to increase

the enrichment limit for fuel storage does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change was evaluated in accordance with the

guidance of the NRC Position Paper entitled, ``OT Position for

Review and Acceptance of Spent Fuel Storage and Handling

Applications'', appropriate sections of the NRC Standard Review

Plan, Regulatory Guides, industry codes, and standards. In addition,

the NRC Safety Evaluation Report for Amendment 155 was also reviewed

with respect to the proposed change.

No new or different mode of operation is proposed. No unproven

technology was utilized in the analytical techniques necessary to

justify the planned fuel storage change. The analytical techniques

used have been developed and used in over 15 applications previously

approved by the NRC. Based upon the reviews, it is concluded that

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

different type accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The only margin of safety potentially impacted by the proposed

change is related to nuclear criticality considerations. The

established acceptance criterion for criticality is that the neutron

multiplication factor in spent fuel pools shall be less than or

equal to 0.95, including all uncertainties, under all conditions.

This margin of safety has been adhered to in the criticality

analysis methods for the proposed change. 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 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: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502

NRC Project Director: William H. Bateman

PECO Energy Co., Public Service Electric and Gas Co., Delmarva

Power and Light Co., and Atlantic City Electric Co., Dockets Nos.

50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2

and 3, York County, Pennsylvania

Date of application for amendments: December 21, 1995

Description of amendment request: The proposed amendments would

modify the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3

Facility Operating Licenses (FOLs) to provide for elimination of

outdated or superseded material regarding, among other things,

environmental monitoring and modifications to the low pressure coolant

injection system, and for making the FOL of Unit 2 consistent with the

FOL of Unit 3.

Basis for proposed no significant hazards consideration

determination:

As required by 10 50.91(a), the licensee has provided its analysis

of the issue of no significant hazards consideration, which is

presented below:

The changes proposed in the Application do not constitute a

Significant Hazards Consideration in that:

i) The proposed changes do not involve a significant increase in

the probability or

[[Page 10397]]

consequences of an accident previously evaluated because the changes

are purely administrative and do not involve any physical changes to

plant SSC [structures, systems, and components]. Therefore, these

changes will not involve a significant increase in the probability

or consequences of an accident previously evaluated.

ii) The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously evaluated

because the changes will not alter the plant or the manner in which

the plant is operated. The changes do not allow plant operation in

any mode that is not already evaluated in the safety analysis. The

changes will not alter assumptions made in the safety analysis and

licensing bases. Therefore, these changes will not create the

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

previously evaluated.

iii) The proposed changes do not involve a significant reduction

in a margin of safety because they are purely administrative and

have no impact on any safety analysis assumptions.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

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

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

Pennsylvania 17105.

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

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

Pennsylvania 19101

NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket No. 50-388,

Susquehanna Steam Electric Station, Unit 2, Luzerne County,

Pennsylvania

Date of amendment request: January 11, 1996

Description of amendment request: The proposed amendment adds a new

action statement to Section 3.8.3.1. of the Technical Specifications

which precludes the need for entry into Limiting Condition for

Operation (LCO) 3.0.3 to allow the performance of certain Emergency

Diesel Generator testing.

Basis for proposed no significant hazards consideration

determination: As required by 10 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:

I. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to allow 8 hours to perform Emergency Diesel

Generator testing and eliminate the need to enter LCO 3.0.3 to

perform this testing does not increase the chances for a previously

analyzed accident to occur. The 8 hour time limit before requiring a

unit shutdown balances the benefit of performing the required test

with the low probability of a LOCA/LOOP [loss-of-coolant accident/

loss of offsite power] while being in the degraded condition for the

duration of the test. To ensure that this risk is minimized, a

significant amount of precautions are taken prior to test

initiation. The governing surveillance procedures have a very

restrictive list of test prerequisites and limitations, which ensure

the availability of remaining ac [alternating current] electrical

power distribution systems and reduce the potential for any single

failure. The allowance of 8 hours to complete the required test

prior to initiating shutdown actions ensures operator attention is

focused on minimizing the potential loss of power to the remaining

division, and restoring power to the effected division upon test

completion; thus, not redirecting operator attention towards a plant

shutdown per 3.0.3. Therefore, the proposed change will not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

II. Create the possibility of a new or different kind of

accident from any accident previously evaluated.

Inhibiting the ESS [electronic switching system] Buses in Unit 1

requires that an LCO be entered in Unit 2 due to the common loads

shared between the Units. However, performance of the LOCA/LOOP or

LOOP surveillance procedures does not cause any diesel generator to

become inoperable as a result of inhibiting an ESS Bus. The time

frame the diesels are fully loaded in the testing evolution is for a

five-minute period to fulfill a Technical Specification requirement.

If at that precise moment a LOCA/LOOP occurs in the operating unit,

the ESS Buses in Unit 1 and 2 will de-energize except for the ESS

Buses that are already connected to the diesels. In the first few

minutes of a postulated LOCA/LOOP occurring in the operating Unit

while performing a LOCA/LOOP test, the operator would have to take

immediate action to shed non-essential loads from the diesels in the

Unit under test to prepare the diesels for the shutdown loads via

the load sequence timers in the operating unit. Existing emergency

procedures require that these actions will be taken. Therefore, the

incorporation of this change will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

With one or more required ac buses, (two load groups) de-

energized, the remaining ac electrical power distribution subsystems

are capable of supporting the minimum safety functions necessary to

shutdown the reactor and maintain it in a safe shutdown condition,

assuming no single failure. The overall reliability is reduced,

however, because a single failure in the remaining power

distribution subsystems could result in the minimum required ESF

[engineered safety feature] functions not being supported.

Therefore, the required ac buses must be restored to OPERABLE status

within a relatively short period of time. Eight hours has been

accepted by the NRC as documented in NUREG-1433, Revision 1,

``Standard Technical Specifications.'' Therefore, the incorporation

of this change will not involve a significant reduction in the

margin to 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)

aresatisfied. 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

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: February 5, 1996

Description of amendment request: The proposed amendment would

revise Davis-Besse Nuclear Power Station (DBNPS) Technical

Specification (TS) 3/4.3.2.1 - Safety Features Actuation System

Instrumentation and its associated Bases. The revision changes the

following items in the Sequence Logic Channels portion of Table 3.3-3:

Functional Unit 4.a, Sequencer; Functional Unit 4.b, Essential Bus

Feeder Breaker Trip (90%); Functional Unit 4.c, Diesel Generator Start,

Load Shed on Essential Bus (59%); and the associated Bases, to clarify

the design and actuation logic and to specify actions to take if

instrumentation channels become inoperable.

Basis for proposed no significant hazards consideration

determination:

As required by 10 50.91(a), the licensee has provided its analysis

of the issue of no significant hazards consideration, which is

presented below:

Toledo Edison has reviewed the proposed changes and determined

that a significant hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power Station, Unit No. 1 in

accordance with these changes would:

1a. Not involve a significant increase in the probability of an

accident previously

[[Page 10398]]

evaluated because the proposed change to accurately reflect the

design and actuation logic of the sequencers and essential bus

undervoltage relays, and provide TS actions for two inoperable

functional units does not make a change to any accident initiator,

initiating condition or assumption. The accident previously

evaluated in the DBNPS Updated Safety Analysis Report (USAR) Section

15.2.9, Loss of All AC Power to the Station Auxiliaries (Station

Blackout), is not affected by this proposed change. The proposed

action statements maintain the USAR requirement for starting and

loading of one [emergency diesel generator] EDG to meet the minimum

[engineered safety features] ESF requirements. The proposed change

accurately reflects the plant design, therefore, the change does not

involve a significant change to the plant design or operation.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because the proposed changes do not

invalidate assumptions used in evaluating the radiological

consequences of an accident, do not alter the source term or

containment isolation and do not provide a new radiation release

path or alter potential radiological releases.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because the proposed

changes do not introduce a new or different accident initiator or

introduce a new or different equipment failure mode or mechanism.

3. Not involve a significant reduction in a margin of safety

because the proposed changes do not reduce the margin to safety

which exists in the present TS or USAR. The proposed changes permit

continued operation with one unit of the sequencer, 59% or 90%

undervoltage protection inoperable provided the unit is placed in

the tripped condition which is consistent with the current TS. With

two units of the same function inoperable the associated EDG is

declared inoperable and the requirements of the TS for an inoperable

EDG entered, including verification that the requirements of TS

3.0.5 are met to assure that the minimum ESF requirement is met. The

operability requirements of the proposed TS are consistent with the

initial condition assumptions of the safety analyses.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

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

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

NRC Project Director: Gail H. Marcus

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: February 5, 1996

Description of amendment request: The proposed amendment would

correct typographical errors, textual inconsistencies, and minor

errors. In addition, equipment identification numbers would be added to

the tables.

Basis for proposed no significant hazards consideration

determination:

As required by 10 50.91(a), the licensee has provided its analysis

of the issue of no significant hazards consideration, which is

presented below:

1. The administrative changes proposed herein will have no

effect on plant hardware, plant design, safety limit setting, or

plant system operation and therefore do not modify or add any

initiating parameters that would significantly increase the

probability or consequences of any previously analyzed accident.

2. These changes do not affect any equipment nor do they involve

any potential initiating events that would create any new or

different kind of accident. Therefore, the proposed change does not

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

any accident previously evaluated.

3. These changes do not affect any equipment involved in

potential initiating events or safety limits. Therefore, it is

concluded that 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

Attorney for licensee: R. K. Gad, III, Ropes and Gray, One

International Place, Boston, MA 02110-2624

NRC Project Director: Ledyard B. Marsh

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

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

Creeks, Manitowoc County, Wisconsin

Date of amendment request: February 8, 1996

Description of amendment request: The proposed amendments will

modify Technical Specification Section 15.3.10, ``Control Rod and Power

Distribution Limits,'' and Section 15.4.1, ``Operational Safety

Review.'' Changes and additions are proposed to clarify the

specifications and to more closely conform to current staff guidance.

Basis for proposed no significant hazards consideration

determination:

As required by 10 50.91(a), the licensee has provided its analysis

of the issue of no significant hazards consideration which is presented

below:

1. Operation of this facility under the proposed Technical

Specifications change will not create a significant increase in the

probability or consequences of an accident previously evaluated.

The probabilities of accidents previously evaluated are based on

the probability of initiating events for these accidents. Initiating

events for accidents previously evaluated for Point Beach include:

control rod withdrawal and drop, CVCS [chemical and volume control

system] malfunction (boron dilution), startup of an inactive reactor

coolant loop, reduction in feedwater enthalpy, excessive load

increase, losses of reactor coolant flow, loss of external

electrical load, loss of normal feedwater, loss of all AC power to

the auxiliaries, turbine overspeed, fuel handling accidents,

accidental releases of waste liquid or gas, steam generator tube

rupture, steam pipe rupture, control rod ejection, and primary

coolant system ruptures.

The consequences of the accidents previously evaluated in the

PBNP [Point Beach Nuclear Plant] FSAR [Final Safety Analysis Report]

are determined by the results of analyses that are based on initial

conditions of the plant, the type of accident, transient response of

the plant, and the operation and failure of equipment and systems.

This change request proposes to improve the clarity of the

requirements concerning shutdown margin, rod group alignment limits,

rod position indication, bank insertion limits, power distribution

limits, at-power physics tests exceptions, and low power physics

tests exceptions. The proposed changes do not affect the probability

of any accident initiating event, because these Technical

Specification requirements do not control any factors that could be

accident initiators. These Technical Specifications establish the

requirements that provide the limitations on the initial conditions,

transient response of the plant, and operation and failure of

equipment and systems. The proposed changes establish the

appropriate limiting conditions for operation, action statements,

and allowable outage times that will continue to ensure that the

results of the accident analyses are not changed. Additionally,

there is no physical change to the facility or its systems.

Therefore, the probability and consequences of any accident

previously evaluated is not increased.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

New or different kinds of accidents can only be created by new

or different accident initiators or sequences. This change request

proposes to improve the clarity of the

[[Page 10399]]

Technical Specifications requirements contained in Technical

Specification Section 15.3.10. The proposed specifications will

clarify the existing Technical Specifications where identified by

rewording, supplementing, or replacing existing requirements. There

is no physical change to the facility or its systems. Therefore, a

new or different kind of accident cannot occur, because no factors

have been introduced that could create a new or different accident

initiator.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are based on the design

and operation of the reactor and containment and the safety systems

that provide their protection.

This change request proposes to improve the clarity of the

Technical Specifications requirements contained in Technical

Specification Section 15.3.10. The proposed specifications will

clarify the existing Technical Specifications where identified by

rewording, supplementing, or replacing existing requirements. There

is no physical change to the facility or its systems. Section

15.3.10 of the Technical Specifications provides the requirements

that limit the operation of the reactor and establish the

operability requirements for reactivity control by the control rod

system. The proposed Technical Specifications changes continue to

provide the appropriate limiting conditions for operation, action

statements, and allowable outage times that ensure the applicable

margins of safety to protect the reactor are preserved. Therefore,

no reduction in any margin of safety has been introduced.

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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

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

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

NRC Project Director: Gail H. Marcus

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

Carolina Power and Light Company, Docket No. 50-400, Shearon Harris

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

Carolina

Date of application for amendments: February 16, 1996

Brief description of amendments: The amendments provide a one-time

surveillance requirement extension for the performance of the trip

actuating device operational test for one of the safety injection

manual initiation switches.

Date of publication of individual notice in Federal Register:

February 26, 1996 (61 FR 7125)

Expiration date of individual notice: March 27, 1996

Local Public Document Room location: Cameron Village Regional

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

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

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

Date of application for amendments: December 19, 1995, as

supplemented by letter dated February 9, 1996.

Brief description of amendments: These amendments allow the

implementation of the recently approved Option B to 10 CFR Part 50,

Appendix J, Option B, by referring to Regulatory Guide 1.163,

``Performance Based Containment Leakage - Test Program.'' This new rule

allows a performance-based option for determining the test frequency

for containment leakage rate testing. The amendment would modify

Technical Specifications (TS) 1.7, 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, and

3/4.6.3, and the Bases of TS 3/4.6.1.2, and would add a new TS 6.16.

Date of issuance: February 23, 1996

Effective date: February 23, 1996, to be implemented within 15 days

of issuance.

Amendment Nos.: Unit 1 - Amendment No. 103; Unit 2 - Amendment No.

92; Unit 3 - Amendment No. 75.

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

amendments revised the Technical Specifications.

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

1627) The February 9, 1996, supplemental letter provided clarifying

information and did not change the initial no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated February 23,

1996.No significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

[[Page 10400]]

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert

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

Date of application for amendment: December 21, 1995

Brief description of amendment: The amendment allows the use of

cladding material other than Zircaloy or ZIRLO. The Safety Evaluation

addresses the safety significance of loading four (4) lead fuel

assemblies (LFAs) into the Calvert Cliffs Nuclear Power Plant, Unit No.

1, reactor vessel during cycles 13, 14, and 15. A Temporary Exemption

was issued on November 28, 1995, (60 FR 62483) approving the loading of

the 4 LFAs into the Unit 1 reactor vessel for the cycles noted above.

The technical basis for the Temporary Exemption, which is the same

basis for the requested TS amendment, was provided in the Baltimore Gas

and Electric Company submittal dated July 13, 1995.

Date of issuance: February 21, 1996

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

30 days.

Amendment No.: 211

Facility Operating License No. DPR-53: Amendment revised the

Technical Specifications.

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

1627) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 21, 1996.No significant hazards

consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

Byron Station, Units 1 and 2, Ogle County, IllinoisDocket Nos. STN

50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will

County, Illinois

Date of application for amendments: June 8, 1995

Brief description of amendments: The amendments revise Technical

Specification (TS) 3/4.8.1 by (1) replacing Table 4.8-1, ``Diesel

Generator Test Schedule,'' with a single surveillance interval of at

least once per 31 days, and (2) deleting TS 4.8.1.1.3, ``Reports.'' The

amendments also revise ACTION statements and surveillances in TS

3.8.1.1 related to certain diesel generator testing and startup

requirements.Date of issuance: February 16, 1996Effective date:

Immediately, to be implemented within 90 days.

Amendment Nos.: 79 and 71

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

The amendments revised the Technical Specifications.

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

45176) The Commission's related evaluation of the amendments is

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

hazards consideration comments received: No

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

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

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

Wilmington, Illinois 60481.

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

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

Date of application for amendments: June 8, 1995

Brief description of amendments: The amendments revise Technical

Specification (TS) 3/4.8.1 by (1) replacing Table 4.8.1.1.2-1, ``Diesel

Generator Test Schedule,'' with a single surveillance interval of at

least once per 31 days, and (2) deleting TS 4.8.1.1.3, ``Reports.'' The

amendments also revise ACTION statements and surveillances in TS

3.8.1.1 related to certain diesel generator testing and startup

requirements.

Date of issuance: February 16, 1996

Effective date: Immediately, to be implemented within 90 days.

Amendment Nos.: 109 and 94

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

revised the Technical Specifications.

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

45176) The Commission's related evaluation of the amendments is

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

hazards consideration comments received: No

Local Public Document Room location: Jacobs Memorial Library,

Illinois alley Community College, Oglesby, Illinois 61348.

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

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

Date of application for amendment: November 3, 1995

Brief description of amendment: This amendment allows deferral of

the Reactor Coolant Pump flywheel inspection until outage 11, scheduled

for the spring of 1998.

Date of issuance: February 15, 1996

Effective date: February 15, 1996

Amendment No.: 153

Facility Operating License No. DPR-72. Amendment revised the

Technical Specifications.

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

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

contained in a Safety Evaluation dated February 15, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

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

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

Date of application for amendments: May 5, 1995, as supplemented by

letter dated September 28, 1995

Brief description of amendments: The amendments consist of changes

to the Technical Specifications (TS) relating to implementation of a

revised thermal design procedure and steam generator water level low-

low setpoint

Date of issuance: February 20, 1996

Effective date: February 20, 1996

Amendment Nos.: 183 and 177Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

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

54719) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 20, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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

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

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

Matagorda County, Texas

Date of amendment request: May 22, 1995, as supplemented by letter

dated October 9, 1995.

Brief description of amendments: The amendments revised Technical

Specification 4.8.1.1.2.e.7 to allow the performance of the 24-hour

surveillance test of the diesel generators during power operation.Date

of issuance: February 21, 1996Effective date: February 21, 1996, to be

implemented within 30 days of issuance.

[[Page 10401]]

Amendment Nos.: Unit 1 - Amendment No. 81; Unit 2 - Amendment No.

70

Facility Operating License Nos. NPF-76 and NPF-80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 19, 1995 (60 FR

37091) The October 9, 1995, supplement provided clarifying information

and did not change the original no significant hazards consideration

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

contained in a Safety Evaluation dated February 21, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Wharton County Junior College,

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

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

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

Illinois

Date of application for amendment: October 27, 1995

Brief description of amendment: The amendment revises Technical

Specification (TS) 3.1.3, ``Control Rod OPERABILITY,'' to include the

25% surveillance overrun allowed by Limiting Condition for Operation

(LCO) 3.0.2 into the allowances of the surveillance Notes for control

rod ``notch'' testing per Surveillance Requirement (SR) 3.1.3.2 and SR

3.1.3.3. The amendment also includes a clarification to the description

of TS Table 3.3.3.1-1, ``Post Accident Monitoring Instrumentation,''

Function 7, to indicate that the Function's requirements apply to the

position indication for only automatic primary containment isolation

valves, rather than all primary containment isolation valves. Finally,

the amendment includes changes to correct a number of editorial and

typographical errors inadvertently contained in TS 3.3.4.1, ``End of

Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation,'' TS 3.3.6.1,

``Primary Containment and Drywell Isolation Instrumentation,'' TS

3.3.8.2, ``Reactor Protection System (RPS) Electric Power Monitoring,''

and TS 3.6.5.2, ``Drywell Air Lock.''

Date of issuance: February 29, 1996

Effective date: February 29, 1996

Amendment No.: 102

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

Technical Specifications.

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

FR 65680) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

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

Atomic Power Station, Lincoln County, Maine

Date of application for amendment: August 30, 1995, as supplemented

by letter dated January 15, 1996.

Brief description of amendment: The amendment revises Technical

Specification 1.3, ``Reactor'', to (1) allow the use of fuel rods clad

with Zircaloy or ZIRLO, rather than restrict use to fuel rods clad with

Zircaloy-4, and (2) replace the specified enrichment limit with a

limitation similar to that found in NUREG-1432, ``Standard Technical

Specifications for Combustion Engineering Plants.''

Date of issuance: February 29, 1996

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

concurrent with Amendment No. 144.

Amendment No.: 155

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

Technical Specifications.

Date of initial notice in Federal Register: October 11, 1995 (60 FR

52932) The January 15, 1996, submittal provided clarifying information

and did not change the initial proposed no significant hazards

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

contained in Safety Evaluation dated February 29, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Wiscasset Public Library, High

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

Niagara Mohawk Power Corporation, Docket Nos. 50-220, and 50-410,

Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County,

New York

Date of application for amendments: October 25, 1995, as

supplemented February 7, 1996.

Brief description of amendments: The amendments revise portions of

Chapter 6 of the Technical Specifications to reflect management

position title and responsibility changes.Date of issuance: February

20, 1996

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

60 days.

Amendment Nos.: 157 and 71

Facility Operating License Nos. DPR-63 and NPF-69: Amendments

revise the Technical Specifications.

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

FR 57605) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 20, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

Pennsylvania

Date of application for amendments: August 1, 1995

Brief description of amendments: These amendments revise the

Technical Specifications Section 3/4.9.1, ``Reactor Mode Switch,'' in

order to provide alternate actions to allow the continuation of core

alterations in the event certain Reactor Manual Control System (RMCS)

and refueling interlocks are inoperable, while preserving the intended

function of the inoperable interlocks.

Date of issuance: February 23, 1996

Effective date: As of date of issuance, to be implemented within 30

days.

Amendment Nos.: 114 and 76

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

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

FR 49944) The Commission's related evaluation of the amendments is

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

hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: October 31, 1994

Brief description of amendment: This amendment deletes certain

valves from Technical Specification Table 3.6.3-1, ``Primary

Containment Isolation Valves,'' that no longer need to be tested in

accordance with 10 CFR Part 50, Appendix J.

Date of issuance: February 22, 1996

[[Page 10402]]

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

60 days from the date of issuance.

Amendment No.: 93

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

Technical Specifications.

Date of initial notice in Federal Register: March 29, 1995 (60 FR

16198) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No

Local Public Document Room location: Pennsville Public Library,

190 S. Broadway, Pennsville, New Jersey 08070

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

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: February 5, 1996, as

supplemented by letter dated February 14, 1996.

Brief description of amendment: The amendment changes Technical

Specifications 4.6.2.2b, ``Suppression Pool Spray,'' and 4.6.2.3b,

``Suppression Pool Cooling,'' to include flow through the RHR heat

exchanger bypass line (in addition to the RHR heat exchanger) in the

Suppression Pool Cooling and Suppression Pool Spray flow path used

during RHR pump testing.

Date of issuance: February 26, 1996

Effective date: As of date of issuance, to be implemented within 3

days.

Amendment No.: 94

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

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes (61 FR 5040) February 9, 1996.

That notice provided an opportunity to submit comments on the

Commission's proposed no significant hazards consideration

determination. No comments have been received. The notice also provided

for an opportunity to request a hearing by March 11, 1996, but

indicated that if the Commission makes a final no significant hazards

consideration determination any such hearing would take place after

issuance of the amendment.The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated February 26, 1996.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

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

County, New Jersey

Date of application for amendments: September 28, 1995

Brief description of amendments: The changes relocate ``Reactor

Coolant System - Chemistry'' Technical Specification 3/4.4.7 for Salem

Unit 1 and 3/4.4.8 for Salem Unit 2 and their associated Bases to the

Salem Updated Final Safety Analysis Report and the Surveillance

Requirements and Limiting Conditions for Operations to applicable plant

procedures controlled by the 10 CFR 50.59 process. Also, the

applicability will be changed from ``At all times'' to ``Modes 1, 2, 3,

4, 5, and 6.''

Date of issuance: February 22, 1996

Effective date: Units 1 and 2, as of date of issuance and shall be

implemented within 60 days of date of issuance.

Amendment Nos.: 180 and 161

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 8, 1995 (60 FR

56369) The Commission's related evaluation of the amendments is

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

hazards consideration comments received: No

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

West Broadway, Salem, New Jersey 08079

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

Seco Nuclear Generating Station, Sacramento County, California

Date of application for amendment: June 20, 1995, as supplemented

on December 19, 1995 and February 7, 1996.

Brief description of amendment: This amendment modifies the

technical specification requirements on qualifications for reviewers of

facility modifications, programs, and documents affecting nuclear

safety and changes the required schedule for reporting changes

requested to environmental permits.

Date of issuance: February 26, 1996

Effective date: February 26, 1996

Amendment No.: 124

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

Technical Specifications.

Date of initial notice in Federal Register: July 19, 1995 (60 FR

37099) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 26, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: Central Library, Government

Documents, 828 I Street, Sacramento, California 95814

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: November 20, 1995

Brief description of amendment: The amendment adds the following

footnote to Technical Specification (TS) 3/4.5.2: ``The allowable

outage time for each RHR train may be extended to 7 days for the

purpose of maintenance and modification. This exception may only be

used one time per RHR train and is not valid after December 31, 1997.''

Date of issuance: February 21, 1996

Effective date: February 21, 1996

Amendment No.: 132

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

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

FR 65684) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 21, 1996. No

significant hazards consideration comments received: No

Local Public Document Room location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180

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

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

San Diego County, California

Date of application for amendments: October 14, 1992, as

supplemented by letter dated December 18, 1995

Brief description of amendments: These amendments revise TS 3/

4.7.5, ``Control Room Emergency Air Cleanup System,'' by reducing the

test duration for the control room emergency air cleanup system and

deleting requirements for duct heaters and diverting valves. The

associated Bases are also revised to reflect these changes.

Date of issuance: February 28, 1996

Effective date: February 28, 1996, to be implemented within 30 days

of issuance.

Amendment Nos.: Unit 1 - Amendment No. 128; Unit 2 - Amendment No.

117

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

revised the Technical Specifications.

[[Page 10403]]

Date of initial notice in Federal Register: March 3, 1993 (58 FR

12267) The December 18, 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 amendments is contained in a Safety Evaluation dated

February 28, 1996.No significant hazards consideration comments

received: No.

Local Public Document Room location: Main Library, University of

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

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

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

Alabama

Date of application for amendments: December 8, 1995 supplemented

January 10, 1996 (TS 364)

Brief description of amendment: The amendments implement recent

changes to 10 CFR 50 Appendix J for performance-based testing of

containment leakage.

Date of issuance: February 22, 1996

Effective Date: February 22, 1996

Amendment Nos.: 228, 243 and 203

Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:

Amendments revised the Technical Specifications.

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

1637) The letter dated January 10, 1996 provided information that did

not change the initial proposed finding of no significant hazards

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

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

hazards consideration comments received: None

Local Public Document Room location: Athens Public library, South

Street, Athens, Alabama 35611

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

Power Company, Toledo Edison Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: November 22, 1995

Brief description of amendment: The amendment added OES Nuclear,

Inc. as an owner.

Date of issuance: February 27, 1996

Effective date: February 27, 1996

Amendment No.: 81

Facility Operating License No. NPF-58: This amendment revised the

license.

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

FR 65685) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 27, 1996. No

significant hazards consideration comments received: No

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of application for amendment: December 12, 1995, supplemented

by facsimile transmission dated January 26, 1996

Brief description of amendment: This amendment revises TS 3/

4.6.1.1, Containment Systems - Primary Containment -Containment

Integrity; TS 3/4.6.1.2, Containment Systems - Containment Leakage; TS

3/4.6.1.6, Containment Systems - Containment Vessel Structural

Integrity; TS 3/4.6.5.3, Containment Systems - Shield Building

Structural Integrity; and associated Bases. The revisions incorporate

changes to the TS to adopt the provisions of Appendix J, Option B for

Type A containment leakage testing as modified by approved exemptions

and in accordance with Regulatory Guide 1.163, to provide consistency

with these new requirements, and to make administrative changes.

Date of issuance: February 22, 1996

Effective date: February 22, 1996, and implemented not later than

90 days after issuance.

Amendment No.: 205

Facility Operating License No. NPF-3. Amendment revised the

Technical Specifications.

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

1637) The January 26, 1996, facsimile transmission was clarifying in

nature and did not affect the initial no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated February 22,

1996.No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

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

October 20, 1995, December 13, 1995, and January 26, 1996.

Brief description of amendment: The amendment revised the allowed

outage time for one unavailable emergency diesel generator from 72

hours to 7 days.

Date of issuance: February 26, 1996

Effective date: February 26, 1996

Amendment No.: 206

Facility Operating License No. NPF-3. Amendment revised the

Technical Specifications.

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

39453) Supplemental information submitted on October 20, 1995, December

13, 1995, and January 26, 1996, provided clarification only and was not

outside the scope of the original no significant hazards determination.

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

Safety Evaluation dated February 26, 1996. No significant hazards

consideration comments received: No

Local Public Document Room location: University of Toledo Library,

Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of application for amendment: September 29, 1995

Brief description of amendment: The amendment increases the minimum

available borated water volume requirement for the boric acid addition

system, the minimum and maximum boron concentration requirements for

the borated water storage tank, the minimum boron concentration

requirement for the core flood tanks; modifies the surveillance

requirements for trisodium phosphate dodecahydrate; and modifies the

refueling boron concentration and the associated Action statement.

Date of issuance: February 27, 1996

Effective date: February 27, 1996

Amendment No.: 207

Facility Operating License No. NPF-3. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 8, 1995 (60 FR

[[Page 10404]]

56371) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 27, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606.

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: November 29, 1994

Brief description of amendments: The amendments revise and update

the North Anna Units 1 and 2 Environmental Protection Plan (EPP) to

reflect current obligations to the Commonwealth of Virginia, revise

portions of the transmission corridor rights-of-way erosion control

program for clarification and to be consistent with the state

regulations, eliminate inconsistencies, and delete obsolete material.

Date of issuance: February 20, 1996

Effective date: February 20, 1996

Amendment Nos.: 197 and 198

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

the Technical Specifications.

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

45188) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 20, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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: October 17, 1995, as

supplemented by facsimile dated February 26, 1996.

Brief description of amendments: The amendments revise the North

Anna Units 1 and 2 Technical Specifications (TS) to allow both of the

containment personnel airlock doors to remain open during refueling

operations, delete License Condition 2.G for Unit 1 and 2.I for Unit 2,

which reference the analyses for limiting doses to control room

operators, and modify the TS Bases to clarify the emergency power

system requirements relative to mitigation of the consequences of a

Fuel Handling Accident.

Date of issuance: February 27, 1996

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

30 days from the date of issuance.

Amendment Nos.: Unit 1 - 198; Unit 2 -179

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

the Technical Specifications and License Conditions.

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

187) The February 26, 1996, facsimile provided clarifying information

that did not change the scope of the October 17, 1995, 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 February 27, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: November 22, 1995, as supplemented by

letter dated February 8, 1996.

Brief description of amendment: This amendment allows the personnel

airlock doors to be open during core alterations and movement of

irradiated fuel in containment. The surveillance requirements for

containment penetrations have also been revised to require that each be

in its ``required condition'' instead of ``closed/isolated condition.''

The Bases section has been updated.

Date of issuance: February 28, 1996

Effective date: February 28, 1996, to be implemented within 30 days

of issuance.

Amendment No.: Amendment No. 95

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

Technical Specifications.

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

FR 65687) The February 8, 1996, supplemental letter provided additional

clarifying information and did not change the original no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

February 28, 1996.No significant hazards consideration comments

received: No.

Local Public Document Room locations: Emporia State University,

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

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

66621

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

Energy Act of 1954, as amended (the Act), and the Commission's rules

and regulations. The Commission has made appropriate findings as

required by the Act and the Commission's rules and regulations in 10

CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

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

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

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

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

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

transcribed as appropriate and the licensee has been informed of the

public comments.

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

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

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

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

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

significant hazards

[[Page 10405]]

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 April 12, 1996, the

licensee may file a request for a hearing with respect to issuance of

the amendment to the subject facility operating license and any person

whose interest may be affected by this proceeding and who wishes to

participate as a party in the proceeding must file a written request

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

hearing and a petition for leave to intervene shall be filed in

accordance with the Commission's ``Rules of Practice for Domestic

Licensing Proceedings'' in 10 CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714 which is available at the

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

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

particular facility involved. If a request for a hearing or petition

for leave to intervene is filed by the above date, the Commission or an

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

the Chairman of the Atomic Safety and Licensing Board Panel, will rule

on the request and/or petition; and the Secretary or the designated

Atomic Safety and Licensing Board will issue a notice of a hearing or

an appropriate order.

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

set forth with particularity the interest of the petitioner in the

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

proceeding. The petition should specifically explain the reasons why

intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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

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

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

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

the Gelman Building, 2120 L Street, NW., Washington, DC, by the above

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

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

Commission by a toll-free telephone call to Western Union at 1-(800)

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

should be given Datagram Identification Number N1023 and the following

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

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

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the 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

[[Page 10406]]

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).

Tennesse Valley Authority, Docket No. 50-390, Watts Bar Nuclear

Plant, Unit No. 1, Rhea County, Tennessee

Date of application for amendment: February 26, 1996

Brief description of amendment: The proposed amendment revises

Technical Specifications (TS) to allow implementation of a proposed

plant modification to preclude inadvertent transfer of the turbine-

driven auxiliary feedwater pump suction from the condensate storage

tank to the emergency raw cooling water system.

Date of issuance: February 28, 1996

Effective date: February 28, 1996

Amendment No.: 1

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

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

February 28, 1996.Public comments requested as to proposed no

significant hazards consideration: No

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Dated at Rockville, Maryland, this 6th day of March 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

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

Reactor Regulation

[Doc. 96-5817 Filed 3-12-96; 8:45 am]

BILLING CODE 7590-01-F